attachment 3 hcgs eal technical bases (clean version) (348 ... · ep-hc-325-201 emergency action...

LR-N17-0002

Attachment 3

HCGS EAL Technical Bases (clean version)

(348 Pages)

HCGS ECG – EAL Technical Bases EP-HC-325-200

HOPE CREEK EVENT CLASSIFICATION GUIDE (ECG) EMERGENCY ACTION LEVEL (EAL) TECHNICAL BASES

TABLE OF CONTENTS

Hope Creek Page 1 of 4 Rev. 0

Hope Creek Generating Station

Event Classification Guide

(ECG)

Emergency Action Level Technical Bases

ECG – EAL Bases Front-Matter Materials:



TABLE OF CONTENTS


EP-HC-325-200 HCGS ECG - EAL Technical Bases Table of Contents EP-HC-325-201 Emergency Action Level (EAL) Technical Basis Introduction EP-HC-325-202 ECG Usage ECG EAL Sections – Bases Information: EP-HC-325-203 EAL Bases for Category R1 - Abnormal Rad Levels / Rad Effluent - Offsite

Rad Conditions EP-HC-325-204 EAL Bases for Category R2 – Abnormal Rad Levels / Rad Effluent -

Irradiated Fuel Events EP-HC-325-205 EAL Bases for Category R3 – Abnormal Rad Levels / Rad Effluent - Area

Radiation Levels EP-HC-325-206 EAL Bases for Category R4 – Abnormal Rad Levels / Rad Effluent - Spent

Fuel Transit & Storage EP-HC-325-207 EAL Bases for Category H1 – Hazards - Security EP-HC-325-208 EAL Bases for Category H2 – Hazards - Seismic Event EP-HC-325-209 EAL Bases for Category H3 – Hazards - Natural or Technological Hazard EP-HC-325-210 EAL Bases for Category H4 - Hazards - Fire EP-HC-325-211 EAL Bases for Category H5 – Hazards - Hazardous Gases EP-HC-325-212 EAL Bases for Category H6 – Hazards - Control Room Evacuation EP-HC-325-213 EAL Bases for Category H7 – Hazards - EC Judgment



TABLE OF CONTENTS


EP-HC-325-214 EAL Bases for Category S1 - System Malfunction - Loss of AC Power EP-HC-325-215 EAL Bases for Category S2 - System Malfunction - Loss of DC Power EP-HC-325-216 EAL Bases for Category S3 - System Malfunction – Loss of Control Room

Indications EP-HC-325-217 EAL Bases for Category S4 - System Malfunction – RCS Activity EP-HC-325-218 EAL Bases for Category S5 - System Malfunction – RCS Leakage EP-HC-325-219 EAL Bases for Category S6 - System Malfunction – RPS Failure EP-HC-325-220 EAL Bases for Category S7 - System Malfunction - Loss of

Communications EP-HC-325-221 EAL Bases for Category S8 - System Malfunction - Hazardous Event

Affecting Safety Systems EP-HC-325-222 EAL Bases for Category F1 - Fission Product Barrier - Fuel EP-HC-325-223 EAL Bases for Category F2 - Fission Product Barriers - RCS EP-HC-325-224 EAL Bases for Category F3 - Fission Product Barriers - Containment EP-HC-325-225 EAL Bases for Category C1 - Cold Shutdown / Refuel System Malfunction

– RPV Level EP-HC-325-226 EAL Bases for Category C2 - Cold Shutdown / Refuel System Malfunction

- Loss of AC Power EP-HC-325-227 EAL Bases for Category C3 - Cold Shutdown / Refuel System Malfunction

- RCS Temperature



TABLE OF CONTENTS


EP-HC-325-228 EAL Bases for Category C4 - Cold Shutdown / Refuel System Malfunction - Loss of DC Power

EP-HC-325-229 EAL Bases for Category C5 - Cold Shutdown / Refuel System Malfunction

- Loss of Communications EP-HC-325-230 EAL Bases for Category C6 - Cold Shutdown / Refuel System Malfunction

- Hazardous Event Affecting Safety Systems ECG – EAL Technical Bases Supporting Documents: EP-HC-325-231 - Use of Fission Product Barrier Table (Tab – Attachment 1) EP-HC-325-232 - EAL Bases Figures / Drawings (Tab – Attachment 2) EP-HC-325-233 - EAL Definitions (Tab – Attachment 3) EP-HC-325-234 - Glossary of Abbreviations & Acronyms (Tab – Attachment 4) EP-HC-325-235 - HCGS-to-NEI 99-01 EAL Cross-reference (Tab – Attachment 5) EP-HC-325-236 – Hope Creek EAL Rad Set-Point Calculation Document (Tab – Attachment 6) EP-HC-325-237 – Hope Creek Safe Operation & Shutdown Rooms/Areas Table R-3 & H-2

Bases (Tab – Attachment 7)


Hope Creek Generating Station Event Classification Guide (ECG)

Emergency Action Level Technical Bases



Introduction

Table of Contents

Section Page

1. Purpose............................................................................................................. 2

2. Emergency Classification Descriptions ............................................................. 2

3. Fission Product Barriers .................................................................................... 4

4. EAL Relationship to EOPs ................................................................................ 5

5. Symptom-Based vs. Event-Based Approach .................................................... 5

6. EAL Organization .............................................................................................. 6

7. Operational Condition Applicability ................................................................... 7

8. Event Classification Guide (ECG) Use .............................................................. 9

9. EAL Technical Bases Organization ................................................................... 17

10. References ........................................................................................................ 19



Introduction

1. PURPOSE

This document provides an explanation and rationale for each Hope Creek Generating Station (HCGS) Emergency Action Level (EAL). It should provide historical documentation for future reference and serve as a training aid. Decision-makers responsible for implementation of the Event Classification Guide (ECG) may use this document as a technical reference in support of EAL interpretation. This information may assist the Emergency Coordinator in making classifications, particularly those involving judgment or multiple events. The information may also be useful in training and for explaining event classifications to offsite officials.

The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

This document is controlled pursuant to 10 CFR 50.54(q).

2. Emergency Classification Descriptions

The NRC and Federal Emergency Management Agency (FEMA) established four emergency classes for fixed nuclear facilities.

An emergency class is used for grouping off-normal nuclear power plant conditions according to their relative radiological seriousness and the time sensitive onsite and offsite actions needed to respond to such conditions.

The four emergency classes are (in order of less severe to most severe):

Unusual Event (UE)

Alert (A)

Site Area Emergency (SAE)

General Emergency (GE)



Introduction

2.1 Unusual Event

Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicates a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

2.2 Alert

Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be small fractions of the EPA Protective Action Guideline exposure levels.

2.3 Site Area Emergency

Events are in progress or have occurred which involve an actual or likely failure of plant functions needed for protection of the public or HOSTILE ACTION that result in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guidelines exposure levels beyond the site boundary.

2.4 General Emergency

Events are in process or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that result in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

3. Fission Product Barriers

Many of the EALs derived from the NEI 99-01 methodology are fission product barrier based. That is, the conditions that define the EALs pertain to the loss or potential loss of one or more of the three fission product barriers. “Loss” and “Potential Loss” signify the relative damage and threat of damage to the barrier. “Loss” means the barrier no longer assures containment of radioactive materials; “Potential Loss” infers an increased probability of barrier loss and decreased certainty of maintaining the barrier.

3.1 Barrier Descriptions

The EAL fission product barriers are:



Introduction

Fuel Clad Barrier (FB): The Fuel Clad barrier consists of the zirconium tubes which house the ceramic uranium oxide pellets along with the end plugs which are welded into each end of the fuel rods comprise the Fuel Clad barrier.

Reactor Coolant System Barrier (RB): The Reactor Coolant System barrier includes the reactor vessel shell, vessel head, CRD housings, vessel nozzles and penetrations, and all primary systems directly connected to the RPV up to the outermost primary containment isolation valves.

Containment (CB): The Containment barrier includes the drywell, the torus, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves.

3.2 Emergency Classification Based on Fission Product Barrier Degradation

The following criteria for event classification relate to fission product barrier loss or potential loss:

ALERT

ANY loss or ANY potential loss of either Fuel Clad or RCS

SITE AREA EMERGENCY

Loss or potential loss of ANY two barriers

GENERAL EMERGENCY

Loss of ANY two barriers and loss or potential loss of the third barrier

Discrete threshold values associated with fission product barrier loss and potential loss are given in ECG-EAL Technical Basis Document, Attachment 1, Use of Fission Product Barrier Table. The bases for the thresholds are discussed in the following ECG sections:

EP-HC-325-222 EAL Bases for Fuel Clad Barrier

EP-HC-325-223 EAL Bases for RCS Barrier

EP-HC-325-224 EAL Bases for Containment Barrier

A point system (described on the Barrier Table) is used to determine fission product barrier emergency classification levels as well as Protective Action Recommendations (PARs) if a General Emergency is declared.



Introduction

4. EAL Relationship to EOPs

Where possible, the EALs have been made consistent with and utilize the conditions defined in the HCGS Emergency Operating Procedures (EOPs). While the symptoms that drive operator actions specified in the EOPs are not indicative of all possible conditions which warrant emergency classification, they define the symptoms, independent of initiating events, for which reactor plant safety and/or fission product barrier integrity are threatened. When these symptoms are clearly representative of one of the NEI Initiating Conditions, they have been utilized as an EAL. This permits rapid classification of emergency situations based on plant conditions without the need for additional evaluation or event diagnosis. Although some of the EALs presented here are based on conditions defined in the EOPs, classification of emergencies using these EALs is not dependent upon EOP entry or execution. The EALs can be utilized independently or in conjunction with the EOPs.

5. Symptom-Based vs. Event-Based Approach

To the extent possible, the EALs are symptom-based; that is, the action level threshold is defined by values of key plant operating parameters that identify emergency or potential emergency conditions. This approach is appropriate because it allows the full scope of variations in the types of events to be classified as emergencies. However, a purely symptom-based approach is not sufficient to address all events for which emergency classification is appropriate. Particular events to which no predetermined symptoms can be ascribed have also been utilized as EALs since they may be indicative of potentially more serious conditions not yet fully realized.

6. EAL Organization

6.1 EAL Groups

The EAL scheme is divided into three broad groups:

EALs applicable under all plant Operational Conditions (OPCONs) – This group would be reviewed by the EAL-user any time emergency classification is considered. This group consists of the ECG Radiological EALs and ECG Hazards EALs.

EALs applicable only under hot OPCONs – This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup or Power Operations OPCONs. This group consists of the ECG System Malfunction EALs and ECG Fission Product Barrier EALs.

EALs applicable only under cold OPCONs – This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown or Refueling OPCONs or when the RPV is defueled. This group consists of the ECG Cold Shutdown / Refuel System Malfunction EALs.



Introduction

The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency.

6.2 EAL Categories and Subcategories

Within each EAL group, EALs are assigned to categories/subcategories. Category titles generally align with the EAL Recognition Categories of NEI 99-01.

Subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. Subcategories are used as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds.



Introduction

The HCGS EAL categories/subcategories and their relationship to NEI Recognition Categories are listed below.

Category EAL Subcategory Title (Proc. #) Group: Any Operating Mode:

R – Abnormal Rad Release / Rad Effluent

1 – Offsite Rad Conditions (203) 2 – Irradiated Fuel Events (204) 3 – Area Radiation Levels (205) 4 - Spent Fuel Transit & Storage (206)

H – Hazards & Other Conditions Affecting Plant Safety

1 – Security (207) 2 – Seismic Event (208) 3 – Natural or Technological Hazard (209) 4 – Fire (210) 5 – Hazardous Gases (211) 6 – Control Room Evacuation (212) 7 – EC Judgment (213)

Group: Hot Conditions:

S – System Malfunction 1 – Loss of AC Power (214) 2 – Loss of DC Power (215) 3 – Loss of Control Room Indications (216) 4 – RCS Activity (217) 5 – RCS Leakage (218) 6 – RPS Failure (219) 7 – Loss of Communications (220) 8 – Hazardous Event Affecting Safety Systems (221)

F – Fission Product Barrier Degradation

None

Group: Cold Conditions:

C – Cold Shutdown / Refuel System Malfunction

1 – RPV Level (225) 2 – Loss of AC Power (226) 3 – RCS Temperature (227) 4 – Loss of DC Power (228) 5 – Loss of Communications (229) 6 – Hazardous Event Affecting Safety Systems (230)

7. Operational Condition Applicability

With the exception of Spent Fuel Transit & Storage (which is not assigned an OPCON), NEI 99-01 assigns one or more operational conditions to each EAL. The Spent Fuel Transit & Storage EAL will be applicable in all OPCONs at Hope Creek Generating Station; as such, OPCON applicability is ALL for the Spent Fuel Transit & Storage EAL.



Introduction

7.1 OPCON Definitions

1 Power Operations

Reactor mode switch is in RUN

2 Startup

The mode switch is in STARTUP/HOT STANDBY

3 Hot Shutdown

The mode switch is in SHUTDOWN and reactor coolant temperature is > 200ºF

4 Cold Shutdown

The mode switch is in SHUTDOWN and reactor coolant temperature is ≤ 200ºF

5 Refueling

The mode switch is in REFUEL or SHUTDOWN and reactor coolant temperature is ≤ 140ºF with fuel in the vessel and vessel head closure bolts less than fully tensioned or head removed

7.2 Added NEI 99-01 condition

DEF Defueled

All reactor fuel removed from RPV. (Full core off load during refueling or extended outage).

(Although Defueled is not a Technical Specification defined OPCON, it corresponds to the Refueling OPCON when all reactor fuel is removed from the RPV.)

7.3 Applicability

The OPCON in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the OPCON that determines whether or not an IC is applicable. If an event or condition occurs, and results in an OPCON change before the emergency is declared, the emergency classification level is still based on the OPCON that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the OPCON at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling OPCONs, even if Hot Shutdown (or a higher OPCON) is entered during the subsequent plant



Introduction

response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown OPCON or higher.

8. Event Classification Guide (ECG) Use

NOTE It is expected the Shift Manager (SM) always serves as the Emergency Coordinator (EC) during the initiating event even if the SM is out of the control room. The Control Room Supervisor (CRS) assumes operational command and control responsibility for the shift crew but not as the EC. The CRS should ensure that the SM is immediately called back to the control room on any conditions that require ECG assessment. Only if the SM is not able (sick or hurt) may the CRS serve as the EC.

8.1. EC Judgment

The EALs described in the ECG are not all inclusive and will not identify each and every condition, parameter or event which could lead to an event classification. The following guidance should be used by the EC:

Initiating Conditions (ICs) are not written as EAL thresholds but do provide a one-line event description which will coincide with a more detailed EAL threshold description.

All ICs/EALs have operating condition (OPCON) applicability which must be satisfied for an IC/EAL to be used for event classification.

IF an IC/EAL has been exceeded,

THEN, CLASSIFY the event IAW the EAL.

IF however, it is clear that the IC/EAL has NOT been exceeded (and will not),

THEN DO NOT CLASSIFY the event.

IF exceeding an IC/EAL threshold is questionable, (unclear),

THEN the EC must use his best judgement, as needed review the EAL Basis document and CLASSIFY the event IAW the most appropriate IC/EAL.



Introduction

In any case,

IF the plant conditions are equivalent to one of the four emergency classes as described in Section 2 of EP-HC-325-101,

THEN CLASSIFY the event based on EC discretion IAW EALs HU 7.1 or HA7.1 or HS 7.1 or HG 7.1 in the EC Judgment Section, EP-HC-325-112 of the ECG.

8.2 Assessment Time

8.2.1 Timeliness

Assessment of an emergency condition should be completed in a timely manner, which is considered to be within 15 minutes of when events are known or should have been known. If an EAL specifies a duration time (e.g., loss of annunciators for 15 minutes or longer), the assessment time runs concurrently with the EAL duration time and is the same length.

8.2.2 Duration Time Exceeded

If an event is recognized or reported and the required duration time is known to have already been exceeded or is unknown, the EAL assessment time is also complete and classification per the subject EAL should be declared without hesitation.

8.2.3 EALs That Require Analysis

A small number of EAL thresholds are related to the results of analyses (e.g. dose assessment, chemistry sampling, leak rate calculations, and/or inspections) that are necessary to ascertain whether a numerical EAL threshold has been exceeded, rather than confirming or verifying an alarm or a received report. In most cases the EAL Basis will identify the analysis necessary and its scope. In these limited cases, the 15-minute emergency classification assessment time starts with the availability of analyses results that show the EAL threshold to be exceeded; this is the time that the information is first available. The NRC expects PSEG Nuclear to initiate and complete these analyses with a reasonable sense of urgency using available on-shift expertise.

8.3 Implementing Actions

The ECG is not a stand-alone document. At times, the ECG will refer the user to other attachments or procedures for accomplishment of specific evolutions such as: Accountability, Recovery, development of PARs, etc. The ECG should be considered an "Implementing Procedure" and used in accordance with the requirements of a Level 2 – Reference Use procedure as defined in HU-AA-104-101, Procedure Use and Adherence. The ECG classification sections allow for judgment and decision making as to whether or not an EAL is exceeded.



Introduction

8.4 Classification

NOTE Comparison of redundant instrumentation, indications, and/or alarms should be used to confirm actual plant conditions.

The primary tools for determining the emergency classification level are the EAL flowcharts or EAL wallcharts. The user of the EAL flowcharts or wallcharts may (but is not required to) consult the EAL Technical Basis in order to obtain additional information concerning the EALs under classification consideration. To use the EAL flowcharts or wallcharts, follow this sequence:

1. ASSESS the event and/or plant conditions and DETERMINE which ECG - EAL Group/Section is most appropriate.

2. REVIEW EAL categories and subcategories on the appropriate flowcharts/wallcharts.

3. If using the ECG – EAL flowcharts, for each applicable subcategory, REVIEW EALs in the subcategory beginning with the lowest emergency classification level to the highest classification level (left to right). ENSURE all pages of a particular subcategory being considered are reviewed.

4. If using the ECG – EAL Wallcharts, for each applicable subcategory, REVIEW EALs in the subcategory beginning with the highest emergency classification level to the lowest classification level (left to right).

5. If in OPCON 1, 2 or 3, also REVIEW the Fission Product Barrier (FPB) Table: a. EXAMINE the FPB categories in the left column of the table. b. SELECT the category that most likely coincides with event conditions. c. REVIEW all thresholds in this category for each fission product barrier. d. For each threshold that is exceeded, IDENTIFY its point value and DETERMINE the

classification level in accordance with the instructions on the Fission Product Barrier Table (or in EAL Technical Bases, Attachment 1).

6. REVIEW the associated EALs as compared to the event and SELECT the highest appropriate emergency classification. If identification of an EAL is questionable refer to paragraph 8.1 above.



Introduction

If there is any doubt with regard to assessment of a particular EAL, the ECG-EAL Technical Bases Document should be reviewed. Words contained in an EAL that appear in uppercase and bold print (e.g., VALID) are defined at the end of the bases for the particular EAL or in ECG – EAL Technical Basis Document, Attachment 3, EP-HC-325-233, EAL Definitions. Words or numbers contained in an EAL that are in bold print but not uppercase are EAL threshold values (e.g., ≥ 15 minutes).

7. If an EAL has been exceeded, equal level EALs or lower level EALs are not required to be separately reported as long as the applicable information is communicated to the NRC using ECG Attachment 5, EP-HC-325-F5, NRC Data Sheet & Completion Reference.

8. When the Shift Manager (SM) is the Emergency Coordinator, the Shift Technical Advisor (STA) is responsible to perform an independent verification of the EAL classification. The STA verification does not alleviate the requirement of the SM to make a timely classification. Should the SM fill the STA role, independent verification of the EAL classification will be delegated to another on-shift SRO, the Independent Assessor.

9. IDENTIFY and IMPLEMENT the referenced ECG form based on the Emergency Classification Level.

Unusual Event Implement EP-HC-325-F1

Alert Implement EP-HC-325-F2

Site Area Emergency Implement EP-HC-325-F3

General Emergency Implement EP-HC-325-F4

Unusual Event (Common Site) Implement EP-HC-325-F24

REFER to EP-HC-325-102, Hope Creek Emergency Classification Description Table, as a guide for correct description wording for entry on the Initial Contact Message Form (ICMF) for all EALs.

10. CONTINUE assessment after classification and attachment initiation, by returning to the EAL flow charts / wallcharts to review EALs that may result in escalation/de-escalation of the emergency level.

8.5 Emergency Short Duration Events

A Short Duration emergency event is a transitory event that meets or exceeds one or more EALs for less than 15 minutes (i.e., action is taken and the plant returned to a condition in which no EAL applies). For a Short Duration event the Control Room Staff is aware of the event and realizes that an EAL had been exceeded.



Introduction

Short Duration events that occur will be assessed and emergency classification made, if appropriate, within 15 minutes of control room indications or the receipt of the information, indicating that an EAL has or had been exceeded. This classification is to be made even if no EALs are currently being exceeded (i.e., actions have been taken to stabilize the Plant such that no EALs currently apply).

Guidance for classifying transient events addresses the period of time of event recognition and classification (15 minutes). However, in cases when EAL declaration criteria may be met momentarily during the normal expected response of the plant, declaration requirements should not be considered to be met when the conditions are a part of the designed plant response, or result from appropriate Operator actions.

8.6 Conditions Discovered After-the-Fact

There may be cases in which a plant condition that exceeded an EAL was not recognized at the time of occurrence but is identified well after the condition has occurred (e.g., as a result of routine log or record review), and the condition no longer exists. In these cases, an emergency should not be declared. Reporting requirements of 10 CFR 50.72 are applicable and the guidance of NUREG-1022, should be applied.

An After-the-Fact event is defined as an event that exceeded an EAL threshold and was not recognized at the time of occurrence but is identified after the condition has occurred (e.g., as a result of a routine log review, record review, post trip review, engineering evaluation) and the condition no longer exists.

For an After-the-Fact event the Control Room Staff was either not aware of the event or did not realize that an EAL was exceeded at the time of the occurrence.

Plant emergency events that are in progress or have occurred with ongoing adverse consequences/effects should not be considered After-the-Fact events and should therefore be classified and declared as an ongoing emergency event.

EMERGENCY CONDITIONS - After-the-Fact events that occur will be assessed and evaluated to ensure that no EAL currently applies. An emergency declaration is NOT required and a non-emergency, One-Hour Report should be initiated in accordance with non-emergency Reportable Acion Levels (RALs) in the ECG.



Introduction

8.7 NRC Communications During An Emergency Guidance

Complete and accurate communications with the NRC Operations Center during emergencies is required and expected. The purpose of notifying the NRC within one-hour of an emergency, is to provide event information when immediate NRC action may be required to protect the public health and safety OR when the NRC needs accurate and timely information to respond to heightened public concern. If the information provided is not accurate or does not contain sufficient detail, then the NRC is hampered in doing their job.

The NRC Data Sheet, along with the Initial Contact Message Form, is the primary vehicle to ensure the NRC is kept informed. General Guidance on completing the event description portion of the NRC Data Sheet is provided in the NRC Data Sheet (ECG Attachment 5).

The NRC notification is to be made immediately after notification of State and local agencies and not later than one hour after the licensee declaration as required by 10 CFR 50.72(a)(3).

An open, continuous communication channel shall be maintained with the NRC Operations Center, during the duration of the event, upon request by the NRC.

8.8 Common Site Events Guidance

Selected EALs in Category H, Hazards (Unusual Event level only) have been designated as “Common Site” events. These events will be annotated with the words “Common Site” within the “Action Required” box on the ECG – EAL flowcharts, just below the OPCON applicability line in the wallcharts and just above the classification level in the EAL Basis document.

The Common Site UE ECG Attachment 24, EP-HC-325-F24, Declaration of “Common Site” UE, will direct the SM to establish agreement on which SM will declare and report the event. Therefore, either Salem or Hope Creek will report Common Site Unusual Events, but not both.

Events classified at an Alert or higher level require plant specific information to be provided to the states of New Jersey and Delaware, the NRC, and to PSEG Emergency Response Facilities and therefore will not be classified as common site events.

8.9 EAL Classification Considerations

Planned evolutions involve preplanning to address the limitations imposed by the condition, the performance of required surveillance testing, and the implementation of specific controls prior to knowingly entering the condition in accordance with the specific requirements of the HCGS Technical Specifications. Activities which cause the site to operate beyond that allowed by the HCGS Technical Specifications, planned or unplanned, may result in an EAL threshold being met or exceeded. Planned evolutions



Introduction

to test, manipulate, repair, perform maintenance or modifications to systems and equipment that result in an EAL value being met or exceeded are not subject to classification and activation requirements as long as the evolution proceeds as planned and is within the operational limitations imposed by the specific operating license. However, these conditions may be subject to the reporting requirements of 10 CFR 50.72 and the Reportable Action Level (RAL) Sections of the ECG should be reviewed.

All classifications are to be based upon VALID indications, reports or conditions. Indications, reports or conditions are considered VALID when they are verified by (1) an instrument channel check, or (2) indications on related or redundant indications, or (3) by direct observation by plant personnel, such that doubt related to the indication’s operability, the condition’s existence, or the report’s accuracy is removed. Implicit in this definition is the need for timely assessment.

Although the majority of the EALs provide very specific thresholds, the Emergency Coordinator must remain alert to events or conditions that lead to the conclusion that exceeding the EAL is IMMINENT. If, in the judgment of the Emergency Coordinator, an IMMINENT situation is at hand, the classification should be made as if the threshold has been exceeded. While this is particularly prudent at the higher emergency classification levels (as the early classification may provide for more effective implementation of protective measures), it is nonetheless applicable to all emergency classification levels.

When multiple simultaneous events occur, the emergency classification level is based on the highest EAL reached. For example, two Alerts remain in the Alert category. Or, an Alert and a Site Area Emergency is a Site Area Emergency. Further guidance is provided in RIS 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events.

Another important aspect of usable EAL guidance is the consideration of what to do when the risk posed by an emergency is clearly decreasing. A combination approach involving recovery from General Emergencies and some Site Area Emergencies and termination from Unusual Events, Alerts, and certain Site Area Emergencies causing no long term plant damage appears to be the best choice. Downgrading to lower emergency classification levels adds notifications but may have merit under certain circumstances. Refer to procedure NC.EP-EP.ZZ-0405, Emergency Termination – Reduction – Recovery, for detailed directions.

The logic used for the Fission Product Barrier EALs reflects the following considerations:

The Fuel Clad Barrier and the RCS Barrier are weighted more heavily than the Primary Containment Barrier. Unusual Events associated with RCS and Fuel Clad Barriers are addressed under EALs in Category S, System Malfunctions.



Introduction

The ability to escalate to higher emergency classification levels as an event deteriorates must be maintained. For example, RCS leakage steadily increasing would represent an increasing risk to public health and safety.

The Primary Containment Barrier should not be declared lost or potentially lost based on exceeding Technical Specification action statement criteria, unless there is an event in progress requiring mitigation by the Primary Containment barrier. When no event is in progress (Loss or Potential Loss of either Fuel Clad and/or RCS barrier) the Primary Containment Barrier status is addressed by Technical Specifications.

9. EAL Technical Bases Organization

EAL technical bases are provided for each EAL according to:

EAL category (R, H, S, F and C)

EAL subcategory

Figures cited in EAL basis discussions are provided in ECG-EAL Technical Basis Document, Attachment 2. EAL defined terms and abbreviations and acronyms are listed in ECG-EAL Technical Basis Document, Attachments 3 and 4, respectively.

For each EAL, the following information is provided:

EAL Category Letter & Title

EAL Subcategory Number & Title

Initiating Condition

Site-specific description of the generic IC given in NEI 99-01.

OPCON Applicability

One or more of the following OPCONs comprise the conditions to which each EAL is applicable: 1 - Power Operations, 2 – Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling, DEF - Defueled.

For Fission Product Barrier Table bases, OPCON applicability is always OPCON 1, 2 and 3. For these EALs, the barrier threat (Loss or Potential Loss) is listed.

EAL# and Classification Level (EAL# & Point Value for Fission Product Barrier Table EAL bases):



Introduction

The EAL number is a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

Category R, H, S and C EALs: (Example: SU1.1)

1. First character (letter) – Corresponds to the EAL category (R, H, S or C)

2. Second character (letter) – Emergency classification level: U for Unusual Event, A for Alert, S for Site Area Emergency, or G for General Emergency

3. Third character (number): Subcategory number within the given category. Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).

4. Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).

Selected EALs in Category H have been designated as “Common Site” events. These events are annotated with the phrase “(Common Site)” immediately following the classification level.

Category F Fission Product Barrier EALs: (Example CB4.P)

1. First and second characters (letters) identify the barrier to which the EAL applies.

FB: Fuel Clad Barrier RB: Reactor Coolant Barrier CB: Containment Barrier

2. Third character (number) – Sequential number beginning with the number one (1) for the first threshold in the barrier loss or potential loss of the Fission Product Barrier Table (Attachment 1)

3. Last character (letter) preceded by a dash (-) designates if EAL is for a potential loss or loss of the barrier in question.

P: Potential Loss L: Loss

EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL wallcharts.



Introduction

Basis

The basis discussion applicable to the EAL taken from NEI 99-01.

Explanation/Discussion/Definitions

Description of the site-specific rationale for the EAL.

EAL Basis Reference(s)

Source documentation from which the EAL is derived. The first reference in each list gives the NEI 99-01 IC and example EAL number. A cross-reference of HCGS EALs and NEI 99-01 ICs/EALs is given in ECG-EAL Technical Basis Document, Attachment 5.

10. REFERENCES

10.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805

10.2 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007

10.3 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73

10.4 10 § CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors

10.5 10 § CFR 50.73 License Event Report System

10.6 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants



Usage

EVENT CLASSIFICATION GUIDE (ECG) USE

NOTE It is expected the Shift Manager (SM) always serves as the Emergency Coordinator (EC) during the initiating event even if the SM is out of the control room. The Control Room Supervisor (CRS) assumes operational command and control responsibility for the shift crew but not as the EC. The CRS should ensure that the SM is immediately called back to the control room on any conditions that require ECG assessment. Only if the SM is not able (sick or hurt) may the CRS serve as the EC.

1. General Considerations When making an emergency classification, the Emergency Coordinator must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated OPCON applicability, notes, and the informing basis information. In the Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds. 1.1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 10.6). When assessing an EAL that specifies a time duration for the off-normal condition, the “clock” for the EAL time duration runs concurrently with the emergency classification process “clock.” 1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator’s operability, the condition’s existence, or the report’s accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration.



Usage

1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Coordinator should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary. 1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CFR 50.72 (ref. 9.4). 1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift). 1.6 Emergency Coordinator Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Coordinator with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Coordinator will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier. Initiating Conditions (ICs) are not written as EAL thresholds but do provide a one-line event description which will coincide with a more detailed EAL threshold description.



Usage

All ICs/EALs have OPCON applicability which must be satisfied for an IC/EAL to be used for event classification.

IF an IC/EAL has been exceeded, THEN CLASSIFY the event IAW the EAL.

IF however, it is clear that the IC/EAL has NOT been exceeded (and will not), THEN DO NOT CLASSIFY the event.

IF exceeding an IC/EAL threshold is questionable, (unclear), THEN the EC must use his best judgement, as needed review the EAL Basis document

and CLASSIFY the event IAW the most appropriate IC/EAL.

In any case, IF the plant conditions are equivalent to one of the four emergency classes as

described in Section 2 of EP-HC-325-101, THEN CLASSIFY the event based on EC discretion IAW EALs HU 7.1 or HA7.1 or HS

7.1 or HG 7.1 in the EC Judgment Section, EP-HC-325-112 of the ECG

2. Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related OPCON applicability and notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process “clock” starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process “clock” started. When assessing an EAL that specifies a time duration for the off-normal condition, the “clock” for the EAL time duration runs concurrently with the emergency classification process “clock.” For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 10.6). 2.1 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

If an Alert EAL and a Site Area Emergency EAL are met, whether at one unit or at two different units, a Site Area Emergency should be declared.

There is no “additive” effect from multiple EALs meeting the same ECL. For example:

If two Alert EALs are met, whether at one unit or at two different units, an Alert should be declared.

Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events (ref. 10.2).



Usage

2.2 Consideration of OPCON Changes During Classification The OPCON in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the OPCON that determines whether or not an IC is applicable. If an event or condition occurs, and results in an OPCON change before the emergency is declared, the emergency classification level is still based on the OPCON that existed at the time that the event or condition was initiated (and not when it was declared). Once a different OPCON is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the OPON at the time of the new event or condition. For events that occur in Cold Shutdown or Refueling, escalation is via EALs that are applicable in the Cold Shutdown or Refueling OPCONs, even if Hot Shutdown (or a higher OPCON) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown OPCON or higher. 2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Coordinator must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Coordinator, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures. 2.4 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 10.2). 2.5 Classification of Short Duration Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration. Examples of such events include an earthquake or a failure of the reactor protection system to automatically scram the reactor followed by a successful manual scram.



Usage

2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration – If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:

An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only.

It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a “grace period” during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Coordinator completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions. 2.7 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 10.3) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 (ref. 10.4) within one hour of the discovery of the



Usage

undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

3. Implementing Actions

The ECG is not a stand-alone document. At times, the ECG will refer the user to other attachments or procedures for accomplishment of specific evolutions such as: Accountability, Recovery, development of PARs, etc. The ECG should be considered an "Implementing Procedure" and used in accordance with the requirements of a Level 2 – Reference Use procedure as defined in HU-AA-104-101, Procedure Use and Adherence. The ECG classification sections allow for judgment and decision making as to whether or not an EAL is exceeded.

4. Classification Process

NOTE Comparison of redundant instrumentation, indications, and/or alarms should be used to confirm actual plant conditions.

4.1 The primary tools for determining the emergency classification level are the EAL wallcharts. The user of the EAL wallcharts may (but is not required to) consult the EAL Technical Bases in order to obtain additional information concerning the EALs under classification consideration. To use the EAL wallcharts, follow this sequence:

4.2 Assess the event and/or plant conditions and determine which EAL Group is most appropriate.

4.3 Review EAL categories and subcategories on the appropriate wallcharts.

4.4 For each applicable subcategory, review EALs in the subcategory beginning with the highest emergency classification level to the lowest classification level (left to right).

4.5 If the HOT conditions wallchart is employed, also review the Fission Product Barrier (FPB) Table as follows: a. Examine the FPB categories in the left column of the table. b. Select the category that most likely coincides with event conditions. c. Review all thresholds in this category for each fission product barrier. d. For each threshold that is exceeded, identify its point value and determine the

classification level in accordance with the instructions on the Fission Product Barrier Table (or in EAL Technical Basis, Attachment 1).

4.6 REVIEW the associated EALs as compared to the event and select the highest appropriate emergency classification. If identification of an EAL is questionable refer to paragraph 1 above.



Usage

4.7 If there is any doubt with regard to assessment of a particular EAL, the ECG EAL Technical Basis Document should be reviewed. Words contained in an EAL that appear in uppercase and bold print (e.g., VALID) are defined at the end of the basis for the EAL.

4.8 If an EAL has been exceeded, equal level EALs or lower level EALs are not required to be seperately reported as long as the applicable information is communicated to the NRC using ECG Attachment 5, EP-HC-325-F5, NRC Data Sheet Completion Reference.

4.9 When the Shift Manager (SM) is the Emergency Coordinator, the Shift Technical Advisor (STA) is responsible to perform an independent verification of the EAL classification. The STA verification does not alleviate the requirement of the SM to make a timely classification. Should the SM fill the STA role, independent verification of the EAL classification will be delegated to another on-shift SRO, the Independent Assessor.

4.10 Identify and implement the referenced ECG form based on the Emergency Classification Level.

Unusual Event Implement EP-HC-325-F1 Alert Implement EP-HC-325-F2 Site Area Emergency Implement EP-HC-325-F3 General Emergency Implement EP-HC-325-F4 Unusual Event (Common Site) Implement EP-HC-325-F24

4.11 Continue assessment after classification and attachment initiation, by returning to the EAL wallcharts to review EALs that may result in escalation/de-escalation of the emergency level.

5. NRC Communications During An Emergency Guidance

5.1 Complete and accurate communications with the NRC Operations Center during emergencies is required and expected. The purpose of notifying the NRC within one-hour of an emergency, is to provide event information when immediate NRC action may be required to protect the public health and safety OR when the NRC needs accurate and timely information to respond to heightened public concern. If the information we provide is not accurate or does not contain sufficient detail, then we hamper the NRC from doing their job.

5.2 The NRC Data Sheet, along with the Initial Contact Message Form, is the primary vehicle to ensure the NRC is kept informed. General Guidance on completing the event description portion of the NRC Data Sheet is provided in the NRC Data Sheet (ECG Attachment 5).



Usage

5.3 The NRC notification is to be made immediately after notification of State and local agencies and not later than one hour after the licensee declaration as required by 10 CFR 50.72(a)(3).

5.4 An open, continuous communication channel shall be maintained with the NRC Operations Center, during the duration of the event, upon request by the NRC.

6. Event Retraction Guidance

Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 10.3).

IF an ENS notification to the NRC was made as directed by the applicable ECG Attachment AND it is later determined that the event or condition is not reportable,

THEN the notification may be retracted as follows:

6.1 OBTAIN both the Operations Shift Manager’s and Shift Manager’s approval of any proposed retractions. Ensure Reg Assurance is consulted prior to approval to retract an Event.

6.2 COMPLETE Page 1 of the ECG Attachment 5, EP-HC-325-F5, NRC Data Sheet Completion Reference, providing a retraction of the original notification. Event Description Section of NRC Data Sheet should explain the rationale for the retraction.

6.3 NOTIFY the NRC Operations Center and NRC Resident Inspector.

6.4 RECORD on the NRC Data Sheet the name of the NRC contact that received the retraction information.

6.5 FORWARD the retraction NRC Data Sheet with the rest of the original attachment of the ECG that was implemented when the original notification was made to the Operations Shift Manager.

7. Common Site Events Guidance

7.1 Selected EALs in Category H (Unusual Event level only) have been designated as “Common Site” events. These events will be annotated with the words “Common Site” just below the OPCON applicability line in the wallcharts and next to the classification level in the EAL Bases document.

7.2 The Common Site UE ECG Attachment 8, EP-HC-325-F8, Declaration of “Common Site” UE, will direct the SM to establish agreement on which SM will declare and report the event. Therefore, either Salem or Hope Creek will report Common Site Unusual Events, but not both.



Usage

7.3 Events classified at an Alert or higher level require plant specific information to be provided to the states of New Jersey and Delaware, the NRC, and to PSEG Emergency Response Facilities and therefore will not be classified as common site events.



EAL# RU1.1

EAL Category: R – Abnormal Rad Levels / Rad Effluent

EAL Subcategory: 1 – Offsite Rad Conditions

Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer

OPCON Applicability: All

EAL# & Classification Level: RU1.1 – UNUSUAL EVENT

EAL:

Reading on ANY Table R-1 effluent radiation monitor > column "UE" for ≥ 60 min. (Notes 1, 2, 3)

Note 1: The Emergency Coordinator should declare the event promptly upon determining that the time limit has been exceeded, or will likely be exceeded.

Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit.

Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes.



EAL# RU1.1

Table R-1 Effluent Monitor Classification Thresholds*

Release Point Monitor GE SAE ALERT UE*

Gas

eous

SPDS – (Total) Offsite Gas

Rad Release

SPDS Point B5097

5.25E+08 µCi/sec

5.25E+07 µCi/sec

5.25E+06 µCi/sec

3.0E+04 µCi/sec

OR OR SUM of: SUM of:

FRVS Vent NG 9RX680 + +

North Plant Vent NG 9RX590 + +

South Plant Vent NG 9RX580 + +

Hardened Torus Vent NG 9RX518

Liqu

id

Liquid Radwaste Discharge 9RX508 ---- ---- ---- 2X the High

Alarm Setpoint

Cooling Tower Blowdown 9RX506 ---- ---- ---- 2X the High

Alarm Setpoint

TB Circ Water Discharge 9RX505 ---- ---- ---- 2X the High

Alarm Setpoint

* For high alarm conditions on offgas pretreatment monitor 9RX621 or 9RX622, refer to EAL SU4.1

Basis: This EAL addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared. Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.



EAL# RU1.1

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. Escalation of the emergency classification level would be via IC RA1.

Explanation/Discussion/Definitions:

The column “UE” gaseous release value in Table R-1 represents two (2) times the calculated ODCM release rate limits.

Instrumentation that may be used to assess this EAL is listed below:

SPDS Point B5097 – Offsite Gas Rad Release

The SPDS point represents the total sum of Rad Gas Releases of the below 4 potential release pathways.

9RX680 (RE-4811A) FRVS Vent Noble Gas

FRVS is normally maintained in a standby condition. Upon FRVS actuation and reactor building isolation, FRVS circulates the reactor building air through HEPA and charcoal filters. Releases are made to the atmosphere via the FRVS Vent Exhaust units.

9RX590 (RE-4873B) North Plant Vent (NPV) Noble Gas

The NPV receives discharge from the gaseous radwaste treatment system (Offgas system), Chem Lab exhaust and solid radwaste area exhaust.

9RX580 (RE-4875B) South Plant (SPV) Vent Noble gas

The SPV receives discharge from the service/radwaste building, reactor building, condensate demineralizer, pipe chase, turbine building mechanical vacuum pumps, gland seal exhaust, and other untreated ventilation sources.

9RX518 Hardened Torus Vent (HTV)

The HTV Rad Gas Release value is a calculated release rate based on HRT Rad detectors 9RX516 (low range) or 9RX517 (high range) and the HTV flow rate. The HTV is not a normal release pathway and is only used when the primary containment is vented per EOP/SAMG guidelines.

The column “UE” liquid release values in Table R-1 represent two (2) times the current High Alarm setpoint based on the current liquid release pathway discharge permit for the specified monitor. Instrumentation that may be used to assess this EAL is listed below:



EAL# RU1.1

9RX508 (RE-4861) Liquid Radwaste Discharge (Upper Range is 5.8E-02 uCi/cc)

The Liquid Radwaste Discharge Line Monitor provides the alarm and automatic termination of liquid radioactive material releases from the liquid rad waste treatment and monitoring system.

9RX506 (RE-8817) Cooling Tower Blowdown (Upper Range is 1.0E-02 uCi/cc)

The Cooling Tower Blowdown Effluent Radiation Monitor monitors radioactivity in the cooling tower blowdown before it is discharged into the Delaware River and warns personnel of an excessive amount of radioactivity (greater than ODCM Limits) being released to the environment. The Cooling-Tower Blowdown Effluent Monitor provides an Alarm function only for releases into the environment.

9RX505 (RE-4557) Turbine Building Circulating Water Dewatering Sump Discharge (Upper Range is 5.8E-02 uCi/cc)

The Turbine Building Circulating Water Dewatering Sump Discharge Monitor provides alarm and automatic termination of liquid radioactive releases from the circulating water dewatering sump. The sump pumps discharge to the circulating water system to the cooling tower basin. The Turbine Building Circulating Water

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AU1 Example EAL #1 and #2 2. HCGS Offsite Dose Calculation Manual (ODCM) Section 3.3.7.11 – Radioactive Gaseous

Effluent Monitoring Instrumentation 3. HCGS Offsite Dose Calculation Manual (ODCM), Section 3.3.7.10 – Radioactive Liquid

Effluent Monitoring Instrumentation 4. Hope Creek Radiological EAL Setpoint Calculation Document (Attachment 6)



EAL# RU1.2



Initiating Condition: Any release of gaseous or liquid radioactivity to the environment greater than 2 times the ODCM for 60 minutes or longer



EAL:

Sample analysis for a gaseous or liquid release indicates a concentration or release rate Table R-2 threshold for ≥ 60 min. (Notes 1, 2)



Table R-2 Effluent Sample Classification Thresholds

Release Point Sample Threshold

Gas

eous

FRVS Vent NG 7.10E-03 Ci/cc

I-131 8.20E-06 Ci/cc

North Plant Vent NG 1.52E-03 Ci/cc

I-131 1.80E-06 Ci/cc

South Plant Vent NG 1.44E-04 Ci/cc

I-131 1.68E-07 Ci/cc Unmonitored Isotopic 2 x ODCM 3/4.11.2

Liqu

id

Liquid Radwaste Discharge Isotopic 2 x ODCM 3/4.11.1

Cooling Tower Blowdown Isotopic 2 x ODCM 3/4.11.1

TB Circ Water Discharge Isotopic 2 x ODCM 3/4.11.1

Unmonitored Isotopic 2 x ODCM 3/4.11.1

Basis:

This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.



EAL# RU1.2

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL. This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.). Escalation of the emergency classification level would be via IC RA1.


Releases in excess of two times the site Offsite Dose Calculation Manual (ODCM) Section 3/4.11.1 or 3/4.11.2 limits that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated dose (which is very low in the UNUSUAL EVENT emergency class) is not the primary concern here; it is the degradation in plant control implied by the fact that the release was not isolated within 60 minutes.

Table R-2 provides calculated radiological release noble gas and iodine sample concentrations that equate to a release that is 2 times the ODCM limits (Section 3/4.11.2) of 500 mRem/year as well as specifying liquid release effluent sample streams 2 times the ODCM limits (Section 3/4.11.1.1, Liquid Effluent Concentrations).

Hope Creek has three gaseous release points for which sample concentration thresholds have been calculated.

FRVS – Filtration Recirculation Vent System NPV – North Plant Vent SPV – South Plant Vent

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AU1 Example EAL #3 2. Off-Site Dose Calculation Manual, Section 3/4.11.1.1 – Liquid Effluents Concentrations 3. Off-Site Dose Calculation Manual, Section 3/4.11.2.1 – Gaseous Effluents Dose Rates 4. Hope Creek Radiological EAL Setpoint Calculation Document (Attachment 6)



EAL# RA1.1



Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE


EAL# & Classification Level: RA1.1 – ALERT

EAL:

In the absence of dose assessment results, reading on ANY Table R-1 effluent radiation monitor > column "ALERT" for ≥ 15 min. (Notes 1, 2, 3, 4)




Note 4 The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.



EAL# RA1.1



Gas

eous


Rad Release

SPDS Point B5097

5.25E+08 µCi/sec

5.25E+07 µCi/sec

5.25E+06 µCi/sec

3.0E+04 µCi/sec






Liqu

id


Alarm Setpoint


Alarm Setpoint


Alarm Setpoint


Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS1.



EAL# RA1.1

Explanation/Discussion/Definitions: This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed either:

10 mRem TEDE

50 mRem CDE Thyroid

The column “ALERT” gaseous effluent release values in Table R-1 correspond to calculated doses of 1% (10% of the SAE thresholds) of the EPA Protective Action Guidelines (TEDE or CDE Thyroid).









The SPV receives discharge from the service/radwaste building, reactor building, condensate demineralizer, pipe chase, turbine building mechanical vacuum pumps, gland seal exhaust, and other untreated ventilation sources





EAL# RA1.1

Definitions:

VALID: An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator’s operability, the condition’s existence, or the report’s accuracy is removed. Implicit in this definition is the need for timely assessment.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AA1 Example EAL #1 2. HCGS Offsite Dose Calculation Manual (ODCM) Section 3.3.7.11 – Radioactive Gaseous




EAL#: RA1.2






EAL:

Dose assessment using actual meteorology indicates doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the MINIMUM EXCLUSION AREA (Note 4)

Note 4: The pre-calculated effluent monitor values presented in EALs RA1.1, RS1.1 and RG1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS1.



EAL#: RA1.2


For the purposes of this EAL, the Site Boundary for HCGS is the MINIMUM EXCLUSION AREA distance; see definition below.

Definitions:

MINIMUM EXCLUSION AREA (MEA): The closest location just beyond the OWNER CONTROLLED AREA where a member of the general public could gain access. For Hope Creek the MEA is 0.56 miles.

OWNER CONTROLLED AREA (OCA): Property owned, maintained and controlled by PSEG Nuclear as part of the Salem & Hope Creek Generating Station complex. For the purpose of emergency classification, area from the PSEG Nuclear access road checkpoint and inward towards the stations is considered the OCA.


EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AA1 Example EAL #2



EAL#: RA1.3






EAL:

Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the MINIMUM EXCLUSION AREA for 60 min. of exposure (Notes 1, 2)



Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS1.



EAL#: RA1.3


Dose assessments based on liquid releases are performed per Offsite Dose Calculation Manual.

Definitions:


EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AA1 Example EAL #3 2. Off-Site Dose Calculation Manual, Section 3/4.11.1 – Liquid Effluents Concentrations



EAL#: RA1.4






EAL:

Field survey results indicate EITHER of the following at or beyond the PROTECTED AREA boundary (Notes 1, 2):

Closed window dose rates > 10 mR/hr expected to continue for ≥ 60 min. Analyses of field survey samples indicate I-131 concentration > 3.85E-08 Ci/cc



Basis:

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release). Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RS1.



EAL#: RA1.4


This EAL addresses a radioactivity release field survey I-131 sample concentration or count rate that would result in a Thyroid CDE dose of greater than 50 mrem for one hour of inhalation at or beyond the PROTECTED AREA boundary. This value exceeds 1% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

The Iodine-131 field survey sample concentration and count rate threshold is based on I-131 dose conversion factors (DCFs) from EPA-400. The thresholds are based on a Thyroid-CDE Dose Rate of 50 mrem/hr for I-131.

For the purposes of this EAL, the PROTECTED AREA boundary is used as it is an easily determined location to obtain a field survey dose rate reading or to obtain a field sample.

Definitions:


PROTECTED AREA (PA): A security controlled area within the OWNER-CONTROLLED AREA (OCA) that is enclosed by the security perimeter fence and monitored by intrusion detection systems. Access to the PA requires proper security clearance and is controlled at the Security Center.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AA1 Example EAL #4 2. Off-Site Dose Calculation Manual, Section 3/4.11.1 – Liquid Effluents Concentrations



EAL#: RS1.1



Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE


EAL# & Classification Level: RS1.1 – SITE AREA EMERGENCY

EAL:

In the absence of dose assessment results, reading on ANY Table R-1 effluent radiation monitor > column "SAE" for ≥ 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Coordinator should declare the event promptly upon determining that the time limit

has been exceeded, or will likely be exceeded.






EAL#: RS1.1



Gas

eous


Rad Release

SPDS Point B5097

5.25E+08 µCi/sec

5.25E+07 µCi/sec

5.25E+06 µCi/sec

3.0E+04 µCi/sec






Liqu

id


Alarm Setpoint


Alarm Setpoint


Alarm Setpoint


Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes. Escalation of the emergency classification level would be via IC RG1.



EAL#: RS1.1


This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed 100 mrem TEDE.

The column “SAE” gaseous effluent release value in Table R-1 corresponds to calculated doses of 10% of the EPA Protective Action Guidelines (TEDE).









The SPV receives discharge from the service/radwaste building, reactor building, condensate demineralizer, pipe chase, turbine building mechanical vacuum pumps, gland seal exhaust, and other untreated ventilation sources. 9RX518 Hardened Torus Vent (HTV)


If dose assessment results are available, EAL RS1.2 would dictate the need for a Site Area Emergency classification due to abnormal radiation effluents.

Definitions:

VALID: An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator’s operability,



EAL#: RS1.1

the condition’s existence, or the report’s accuracy is removed. Implicit in this definition is the need for timely assessment.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AS1 Example EAL #1 2. HCGS Offsite Dose Calculation Manual (ODCM) Section 3.3.7.11 – Radioactive Gaseous




EAL#: RS1.2






EAL:

Dose assessment using actual meteorology indicates doses > 100 mrem TEDE or 500 mrem thyroid CDE at or beyond the MINIMUM EXCLUSION AREA (Note 4)


Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC RG1.



EAL#: RS1.2



Definitions:




EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AS1 Example EAL #2 2. Hope Creek Radiological EAL Setpoint Calculation Document (Attachment 6)



EAL#: RS1.3






EAL:

Field survey results indicate EITHER of the following at or beyond the PROTECTED AREA boundary:

Closed window dose rates > 100 mR/hr expected to continue for ≥ 60 min. Analyses of field survey samples indicate I-131 concentration > 3.85E-07 Ci/cc

(Notes 1, 2)



Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Escalation of the emergency classification level would be via IC RG1.


This EAL addresses a radioactivity release field survey I-131 sample concentration or count rate that would result in a Thyroid CDE dose of greater than 500 mrem for one hour of inhalation at or beyond the PROTECTED AREA boundary. This value exceeds 10% of the



EAL#: RS1.3

EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

The Iodine-131 field survey sample concentration and count rate threshold is based on I-131 dose conversion factors (DCFs) from EPA-400. The thresholds are based on a Thyroid-CDE Dose Rate of 500 mrem/hr for I-131.


Definitions:



EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AS1 Example EAL #3 2. HCGS Offsite Dose Calculation Manual Figure 5.1.1-1, Area Plot Plan of Site 3. Hope Creek Radiological EAL Setpoint Calculation Document (Attachment 6)



EAL#: RG1.1



Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE


EAL# & Classification Level: RG1.1 – GENERAL EMERGENCY

EAL:

In the absence of dose assessment results, reading on ANY Table R-1 effluent radiation monitor > column "GE" for ≥ 15 min. (Notes 1, 2, 3, 4) Note 1: The Emergency Coordinator should declare the event promptly upon determining that the time limit







EAL#: RG1.1



Gas

eous


Rad Release

SPDS Point B5097

5.25E+08 µCi/sec

5.25E+07 µCi/sec

5.25E+06 µCi/sec

3.0E+04 µCi/sec






Liqu

id


Alarm Setpoint


Alarm Setpoint


Alarm Setpoint


Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.



EAL#: RG1.1


This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to site boundary doses that exceed 1,000 mrem TEDE.

The column GE gaseous effluent release value in Table R-1 corresponds to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE).









The SPV receives discharge from the service/radwaste building, reactor building, condensate demineralizer, pipe chase, turbine building mechanical vacuum pumps, gland seal exhaust, and other untreated ventilation sources.



If dose assessment results are available, EAL RG1.2 would dictate the need for a General Emergency classification due to abnormal radiation effluents.

Definitions:

VALID: An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by



EAL#: RG1.1

direct observation by plant personnel, such that doubt related to the indicator’s operability, the condition’s existence, or the report’s accuracy is removed. Implicit in this definition is the need for timely assessment.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AG1 Example EAL #1 2. HCGS Offsite Dose Calculation Manual (ODCM) Section 3.3.7.11 – Radioactive Gaseous




EAL#: RG1.2






EAL:

Dose assessment using actual meteorology indicates doses > 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond the MINIMUM EXCLUSION AREA (Notes 3, 4)



Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.



EAL#: RG1.2



Definitions:




EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AG1 Example EAL #2 2. Hope Creek Radiological EAL Setpoint Calculation Document (Attachment 6)



EAL#: RG1.3





EAL# & Classification Level: RG1.3 – GENERAL AREA EMERGENCY

EAL:

Field survey results indicate EITHER of the following at or beyond the PROTECTED AREA boundary:

Closed window dose rates > 1,000 mR/hr expected to continue for ≥ 60 min. Analyses of field survey samples indicate I-131 concentration > 3.85E-06 Ci/cc

(Notes 1, 2)



Basis:

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public. Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.


This EAL addresses a radioactivity release field survey I-131 sample concentration or count rate that would result in a Thyroid CDE dose of greater than 5,000 mRem for one hour of inhalation at or beyond the PROTECTED AREA boundary. This value exceeds 100% of the



EAL#: RG1.3

EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

The Iodine-131 field survey sample concentration and count rate threshold is based on I-131 dose conversion factors (DCFs) from EPA-400. The thresholds are based on a Thyroid-CDE Dose Rate of 5,000 mrem/hr for I-131.


Definitions:



EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AG1 Example EAL #3 2. HCGS Offsite Dose Calculation Manual Figure 5.1.1-1, Area Plot Plan of Site 3. Hope Creek Radiological EAL Setpoint Calculation Document (Attachment 6)



EAL#: RU2.1


EAL Subcategory: 2 – Irradiated Fuel Events

Initiating Condition: UNPLANNED loss of water level above irradiated fuel



EAL:

UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

Confirmed SFP low level alarm Annunciator D1-A5 (FUEL POOL LEVEL HI/LO)

Reactor Water Level Shutdown Range Indicator LI-R605-B21

Visual observation (local or remote)

AND

UNPLANNED rise in corresponding area radiation levels on ANY of the following:

Spent Fuel Storage Pool Area (9RX707)

New Fuel Criticality A Rad (9RX612)

New Fuel Criticality B Rad (9RX613)

Temporary Refueling Bridge ARM

Basis:

This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant. A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnel (e.g., from a refueling crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations. The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor



EAL#: RU2.1

vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an unplanned loss of water level. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level would be via IC RA2.


The fuel pool low level alarm is actuated by level switch LSHL-4657 when fuel pool water level drops below 39 ft 9 in. This alarm is actuated by local panel alarm 10C214 1-1 Fuel Pool Lo Level or 10C214 1-2 Reactor Cavity Low Level. The reactor cavity low level alarms at 75 in. from LISL-N027 (L-11683) reactor vessel level Shutdown Range transmitter 1BBLT-N027 (1CCLT-11683). The annunciator on local Panel 10C214 is always valid for low Spent Fuel Pool level, however it is only enabled for the low reactor cavity water level during refueling operations. The annunciator on local Panel 10C214 alarms on low Spent Fuel Pool level and/or low reactor cavity water level when the annunciator is enabled.

During refueling operations, the reactor cavity is flooded and reactor cavity level indication is monitored on the shutdown instrument range (LI-R605-B21).

When irradiated fuel is not seated in the RPV or SFP storage racks, there still remains the possibility of uncovering irradiated fuel during refueling operations. Therefore, this EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool as well as for spent fuel pool drain down events.

Area radiation monitors that may respond to a loss of spent fuel shielding are those located on the 201' elevation (refuel floor):

Spent Fuel Storage Pool Area (9RX707) New Fuel Criticality A Rad (9RX612) New Fuel Criticality B Rad (9RX613) Temporary Refueling Bridge ARM

Definitions:

REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway

UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.



EAL#: RU2.1

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AU2 Example EAL #1 2. HC.OP-AR.ZZ-0013(Q) Overhead Annunciator Window Box D1-A5 3. HC.OP-AR.EC-0002(Q) Reactor Level Indication Panel 10C214 4. HC.OP-AB.COOL-004(Q) Fuel Pool Cooling 5. HC.OP-AB.CONT-005(Q) Irradiated Fuel Damage 6. UFSAR Table 12.3-9 Area Radiation Monitor Detector Locations 7. UFSAR Section 9.1.3.5 Instrumentation Applications 8. HC.IC-GP.BB-0003(Q) Channel L-11683 / B21-N027 Rx Cavity Floodup Level/RX

Shutdown Range Level Setup



EAL#: RU2.1

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EAL#: RA2.1


EAL Subcategory: 2 – Irradiated Fuel Event

Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel



EAL:

Uncovery of irradiated fuel in the REFUELING PATHWAY

Basis: This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used (e.g., a boil-off curve). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss. A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.

Escalation of the emergency classification level would be via IC RS1.


Definitions:



EAL#: RA2.1

REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway




EAL#: RA2.2






EAL:

Damage to irradiated fuel resulting in a release of radioactivity that causes a High alarm on ANY of the following radiation monitors:

Spent Fuel Storage Pool Area (9RX707)

New Fuel Criticality A Rad (9RX612)

New Fuel Criticality B Rad (9RX613)

Refuel Floor Exhaust Duct Rad Channel A (9RX627)

Refuel Floor Exhaust Duct Rad Channel B (9RX628)

Refuel Floor Exhaust Duct Rad Channel C (9RX629)

Basis:

This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. Escalation of the emergency would be based on either Recognition Category R or C ICs. This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident). Escalation of the emergency classification level would be via IC RS1.


Area radiation monitors that may alarm in response to damaged spent fuel are those located on the 201' elevation (refuel floor):



EAL#: RA2.2

Spent Fuel Storage Pool Area (9RX707) New Fuel Criticality A Rad (9RX612) New Fuel Criticality B Rad (9RX613)

The Refuel Floor Exhaust Duct Rad Monitors 9RX627, 9RX628 and 9RX629 are process monitors and are designed to detect a release of Fission Products to the Reactor Building atmosphere. Hence, they are included as part of the EAL threshold, to confirm the magnitude of damage to an irradiated fuel bundle. These monitors can also react as Area Radiation Monitors, in the event of rising radiation levels due to lowered shielding, as would occur during a loss of spent fuel pool inventory event. It is important to distinguish between the cause for rising radiation levels when classifying an event under this EAL.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AA2 Example EAL #2 2. UFSAR Table 12.3-9 Area Radiation Monitor Detector Locations 3. HC.OP-AR.ZZ-0013(Q) Overhead Annunciator Window Box D1-A5 4. HC.OP-AB.COOL-004(Q) Fuel Pool Cooling 5. HC.OP-AB.CONT-005(Q) Irradiated Fuel Damage 6. HC.RP-AR.SP-0001(Q) RMS Alarm Response 7. HC.OP-AR.ZZ-0019(Q) Overhead Annunciator Window Box E6-A3



EAL#: RA2.3






EAL:

Lowering of spent fuel pool level to 186 ft.

Basis:

This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant. Escalation of the emergency would be based on either Recognition Category R or C ICs. Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool. Escalation of the emergency classification level would be via IC RS1.


Post-Fukushima order EA-12-051 (ref. 2) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). 185 ft. 6 inches (rounded to 186 ft.) - Level 2 corresponds to 10' above top of fuel rack. This level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck.




EAL#: RA2.3

2. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation

3. HC.OP-AR.ZZ-0013(Q) Overhead Annunciator Window Box D1



EAL#: RS2.1



Initiating Condition: Spent fuel pool level at the top of the fuel racks



EAL:

Lowering of spent fuel pool level to 176 ft.

Basis:

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC RG1 or RG2.


Post-Fukushima order EA-12-051 (ref. 2) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). 175.67 ft. (rounded to 176 ft.) - Level 3 corresponds to the top of the fuel rack and the level where the instrument channel is offscale low.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AS2 Example EAL #1 2. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel

Pool Instrumentation 3. HC.OP-AR.ZZ-0013(Q) Overhead Annunciator Window Box D1



EAL#: RS2.1




EAL#: RG2.1



Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the fuel racks for 60 minutes or longer



EAL:

Spent fuel pool level CANNOT be restored to at least 176 ft. for ≥ 60 min. (Note 1) Note 1: The Emergency Coordinator should declare the event promptly upon determining that the time limit


Basis:

This EAL addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncover of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.


Post-Fukushima order EA-12-051 (ref. 2) required the installation of reliable SFP level indication capable of identifying normal level (Level 1), SFP level 10 ft. above the top of the fuel racks (Level 2) and SFP level at the top of the fuel racks (Level 3). 175.67 ft. (rounded to 176 ft.) - Level 3 corresponds to the top of the fuel rack and the level where the instrument channel is offscale low.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AG2 Example EAL #1 2. NRC EA-12-51 Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel

Pool Instrumentation 3. HC.OP-AR.ZZ-0013(Q) Overhead Annunciator Window Box D1



EAL#: RG2.1




EAL#: RA3.1


EAL Subcategory: 3 – Area Radiation Levels

Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown



EAL:

Dose rates > 15 mR/hr in the Control Room (9RX710 or by survey)

Basis:

This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable. Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.


The area that meets this threshold at Hope Creek is the Control Room. The Central Alarm Station (CAS) is not included in this EAL because it is located at Salem. Control Room ARM (9RX710) measures area radiation in a range of 1 - 104 mR/hr.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AA3 Example EAL #1 2. HCGS UFSAR Table 12.3-9 Area Radiation Monitor Detector Locations 3. HC.RP-AR.SP-0001(Q) RMS Alarm Response



EAL#: RA3.1




EAL#: RA3.2


EAL Subcategory: 3 – Area Radiation Levels

Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown

OPCON Applicability: 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling


EAL:

An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to ANY Table R-3 rooms or areas (Note 5)

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then NO emergency classification is warranted.

Table R-3 Safe Operation & Shutdown Rooms/Areas Bldg. – Ele. Rooms/Areas OPCON

Applicability

RB 54’ 4113/4109 (RHR A/B Pump Rooms) 3, 4, 5

RB 77’ 4201 (10B242 MCC) 3, 4, 5

RB 102’ 4307 (B SACS Pump Room) 3, 4, 5

RB 102’ 4328/4322 (North/South HCU) 3, 4, 5

Basis:

This EAL addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or IMPEDE personnel from performing actions necessary to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Coordinator should consider the cause of the increased radiation levels and determine if another IC may be applicable. For RA3.2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated



EAL#: RA3.2

radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits). An emergency declaration is not warranted if any of the following conditions apply:

The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).

The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.


If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

Definitions:

IMPEDE(D): Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).



EAL#: RA3.2


NOTE: EAL RA3.2 mode applicability has been limited to the applicable modes identified in Table R-3 Safe Operation & Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table R-3 are changed, a corresponding change to Attachment 3 ‘Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases’ and to EAL RA3.2 mode applicability is required.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, AA3 Example EAL #2 2. Attachment 7 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases



EAL#: RA3.2




EAL#: RU4.1


EAL Sub-category: 4 - Spent Fuel Transit & Storage

Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY



EAL:

Damage to a Multi Purpose Canister (MPC) CONFINEMENT BOUNDARY as indicated by on-contact radiation readings ≥ 600 mR/hr (gamma + neutron) on the surface of the spent fuel cask, excluding the air vents, OR > 60 mR/hr (gamma + neutron) on the top of the spent fuel cask.

Basis:

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of “damage” is determined by radiological survey. The technical specification multiple of “2 times (ISFSI Technical Specification limits)”, which is also used in Category R IC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the “on-contact” dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSIs are covered under ICs HU1 and HA1.


This EAL applies to emergency conditions affecting a spent fuel cask caused by an accident or natural phenomena. This EAL would be applicable at all times in all Operational Conditions for a loaded spent fuel storage cask from the time the lid is installed, as the cask leaves the Hope Creek Reactor Building and during transport to and



EAL#: RU4.1

storage at the INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI). This EAL provides for an Unusual Event classification, which may be entered if conditions occur that have the potential for damaging or degrading the CONFINEMENT BOUNDARY of a spent fuel cask. Damage to the storage cask could result in an increase in direct radiation readings from the cask. A similar Salem EAL is only applicable for a Salem spent fuel cask that is in transit to the ISFSI. After the spent fuel cask is in place at the ISFSI, any further conditions that could adversely impact the ISFSI or an individual cask from either Salem or Hope Creek would be assessed and classified as needed by the Hope Creek Shift Manager (SM) per this EAL.

As provided in the Holtec HI-STORM 100 System Certificate of Compliance (CoC), Appendix A (Technical Specifications), Section 5.7.4 contains radiation values for the cask that should not be exceeded. Under Amendment #5, the highest allowable radiation level on contact with the HI-STORM 100 cask body is 300 mR/hr on the side of the cask and 30 mR/hr on the top of the cask. Keeping in line with NEI guidance that a UE is warranted for radiation conditions at a level of twice the Technical Specification value, 600 mR/hr and 60 mR/hr are being used as the EAL threshold radiation levels.

Continued use of this lower value is conservative for casks loaded under later CoC amendments where the radiation limit values may increase. The threshold values are sufficiently above nominal radiation levels of the CONFINEMENT BOUNDARY that radiation levels above this EAL threshold would indicate significant damage to the CONFINEMENT BOUNDARY.

No releases of radioactive material requiring offsite response or monitoring are expected because the seal-welded spent fuel canister (part of the CONFINEMENT BOUNDARY) is designed to remain intact under all normal, off-normal, and credible accident conditions of onsite transport and storage at the ISFSI, according to Holtec licensing documents. Prior to the installation of the spent fuel cask lid on the HI-STORM 100 cask, emergency classifications would be based on other Category R EALs.

Definitions:

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

CONFINEMENT BOUNDARY: Is the barrier(s) between areas containing radioactive substances and the environment and includes the multi-purpose canister (MPC) and, for the purposes of this EAL, the associated cask shielding.

EAL Basis Reference(s): 1. NEI 99-01 Rev. 6, E-HU1 Example EAL #1 2. HOLTEC HI-STORM 100 UFSAR, Chapter 5 and Chapter 11 3. Certificate of Compliance, Docket # 72-1014



EAL#: RU4.1

4. Holtec International Final Safety Analysis Report for the HI-STORM 100 Cask System Holtec Report No.: HI-2002444

5. Certificate of Compliance No. 72-1014 Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 1.1 Definitions



EAL#: RU4.1




EAL#: HU1.1

EAL Category: H – Hazards

EAL Subcategory: 1 – Security

Initiating Condition: Confirmed SECURITY CONDITION or threat


EAL# & Classification Level: HU1.1 – UNUSUAL EVENT (Common Site)

EAL:

A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Shift Manager or designee (Note 6)

OR

Notification of a credible security threat directed at the site – (determined by security in accordance with SY-AA-101-132, “Threat Assessment”) (Note 6)

OR

A VALIDATED notification from the NRC providing information of an aircraft threat (Note 6)

Note 6: Shift Manager (SM) should implement the Prompt Actions of the Security Emergency Guideline Attachment located in NC.EP-EP.ZZ-0102, EC Response, prior to classification of a security emergency.

Key Information to obtain from Security Supervision upon SM notification of a security event:

Determination if the security event is a HOSTILE ACTION or SECURITY CONDITION

If a HOSTILE ACTION, is location the OCA or PA?

Basis:

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1, HS1 and HG1. Timely and accurate communications between Security Shift Manager and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations. Security plans and terminology are based on the guidance provided by NEI 03-12, Template



EAL#: HU1.1

for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program. The first threshold references the Security Shift Manager because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information. The second threshold addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the Salem – Hope Creek Security Contingency Plan (ref. 2). The third threshold addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with the Salem – Hope Creek Security Contingency Plan (ref. 2). Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Salem – Hope Creek Security Contingency Plan (ref. 2). Escalation of the emergency classification level would be via IC HA1.


If the security events do not meet the threshold for an UNUSUAL EVENT classification, they may result in the need to make a non-emergency report per RAL Section 11.7.1.a, One Hour Non-Emergency Safeguards Event (10 CFR 73.71) as determined by Security per SY-AA-1002, “Safeguards Event Report.”

Security will be focused on actions to mitigate the security event and will provide the SM with key information as the event progresses. Communications between the SMs and the Security Shift Manager should be accurate, concise, and focused on EAL criteria and protection of key target sets. As Security and Operations terminology sometimes differ, clarifying questions should be asked to ensure accurate information exchange.

1st Condition (SECURITY CONDITION)

Attached to this EAL Basis is a “Security Contingency Event Summary Table” that indicates which Security Contingency Events could result in Security Supervision determining that a SECURITY CONDITION exists and therefore an UNUSUAL EVENT classification should be made OR, could result in Security Supervision determining that a HOSTILE ACTION is or has occurred and therefore classification at the ALERT or higher level should be made based on the location (OCA or PA) of the HOSTILE ACTION.



EAL#: HU1.1

2nd Condition (credible security threat)

This threshold is included to ensure that threat information from any source which is assessed by security supervision as being a “credible security threat” is classified as an UNUSUAL EVENT. For Security Events, Salem and Hope Creek is considered a single site, therefore a “credible security threat” to either Salem or Hope Creek would affect the entire site and a “Common Site” UE declaration would be made.

Timely classification will ensure that Offsite Response Organizations and plant personnel are notified in a timely manner resulting in a state of heightened awareness. Threats are evaluated by security per Threat Assessment, SY-AA-101-132. Security threats that do not meet the definition of a “credible security threat” should be dispositioned IAW Threat Assessment, SY-AA-101-132.

3rd Condition (aircraft threat)

Aircraft threat calls from the NRC should be validated by use of NRC authentication code or a return call to the NRC Headquarter Operations Center.

For security events, Salem and Hope Creek is considered a single site, therefore, a validated aircraft threat to either Salem or Hope Creek would affect the entire site and a “Common Site” UE declaration would be made.

Definitions:

HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION: An act toward Salem or Hope Creek or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Salem or Hope Creek. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OCA).

OWNER CONTROLLED AREA (OCA): Property owned, maintained and controlled by PSEG Nuclear as part of the Salem & Hope Creek Generating Station complex. For the purpose of emergency classification, the area from the PSEG Nuclear access road checkpoint and inward towards the stations is considered the OCA.

PROJECTILE: An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA (PA): A security controlled area within the OWNER-CONTROLLED AREA (OCA) that is enclosed by the security perimeter fence and monitored by intrusion



EAL#: HU1.1

detection systems. Access to the PA requires proper security clearance and is controlled at the Security center.

SECURITY CONDITION: Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.

VALIDATED: As related to an aircraft threat, a call from the NRC that is confirmed to be authentic. Calls from the NRC are VALIDATED by use of the NRC provided authentication code or by making a return call to the NRC Headquarter Operations Center and confirming threat information with the NRC Operation Officer. Aircraft threat calls from other agencies, NORAD, FAA, or FBI should be VALIDATED by calling the NRC Operations Officer.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HU1 Example EAL #1, #2, #3 2. Salem – Hope Creek Security Contingency Plan 3. SY-AA-101-132 Threat Assessment 4. HC.OP-AB.SEC-0001 Security Event 5. HC.OP-AB.SEC-0002 Airborne Threat



EAL#: HU1.1

Security Contingency Event Summary Table

Note: Do not include titles of Security Contingency Event documents in future revisions

Contingency Event

Number

Event Could Result in Determination

of a SECURITY CONDITION

(UE ONLY) Yes / No


of a HOSTILE ACTION (ALERT or Higher)

Yes / No # 1 Yes Yes # 2 Yes Yes # 3 Yes No # 4 No Yes # 5 Yes Yes # 6 No No # 7 No No # 8 Yes Yes # 9 Yes Yes # 10 Yes Yes # 11 Yes No # 12 Yes No # 13 Yes No # 14 Yes No # 15 Yes No # 16 Yes Yes # 17 Yes Yes # 18 No Yes # 19 Yes Yes # 20 Yes No



EAL#: HU1.1




EAL#: HA1.1



Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes


EAL# & Classification Level: HA1.1 – ALERT

EAL:

A HOSTILE ACTION is occurring or has occurred within the OCA as reported by the Security Shift Manager or designee (Note 6)

OR

A VALIDATED notification from NRC of an aircraft attack threat within 30 min. of the site (Note 6)





Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications between the Security Shift Manager and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response



EAL#: HA1.1

Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. The first threshold is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against an ISFSI that is located outside the plant PROTECTED AREA. The second threshold addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with site-specific security procedures (ref. 3, 4). The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC. In some cases, it may not be readily apparent if an aircraft impact within the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Salem – Hope Creek Security Contingency (ref. 2).


This event will be escalated to a SITE AREA EMERGENCY based upon HOSTILE ACTION affecting the PROTECTED AREA (PA). Also, if Salem declares an SAE due to their PA being affected by the security event, Hope Creek will escalate to SAE to match them.

1st Condition (OCA HOSTILE ACTION)

Reference is made to the specific security shift supervision (Security Shift Manager or designee) because these individuals are the designated personnel on-site qualified and trained to confirm that a HOSTILE ACTION is occurring or has occurred.



EAL#: HA1.1

This EAL condition is not premised solely on adverse health effects caused by a radiological release. Rather the issue is the immediate need for assistance due to the nature of the event and the potential for significant and indeterminate damage. Although nuclear plant security officers are well trained and prepared to protect against HOSTILE ACTION, it is appropriate for Offsite Response Organizations (OROs) to be notified and encouraged to begin activation to be better prepared should it be necessary to consider further actions.

Attached to this EAL Basis is a “Security Contingency Event Summary Table” that indicates which Security Contingency Events could result in Security Supervision determining that a HOSTILE ACTION is or has occurred and therefore classification at the ALERT or higher level should be made based on the location (OCA or PA) of the HOSTILE ACTION. Security events that do not involve a HOSTILE ACTION may result in Security Supervision determining that a SECURITY CONDITION exists and therefore an UNUSUAL EVENT classification should be made per EAL HU1.1.

2nd Condition (aircraft attack)

The fact that the site is an identified attack candidate with minimal time available for further preparation requires a heightened state of readiness and implementation of protective measures that can be effective (onsite evacuation, dispersal, or sheltering) before arrival or impact.

This EAL is met when a plant receives validated information regarding an aircraft attack threat from NRC and the aircraft is less than 30 minutes away from the site. Only the site (Salem and Hope Creek is considered a single site for Security event classifications) to which the specific threat is made needs declare the ALERT.

Aircraft threat calls from the NRC should be validated by use of NRC authentication code or a return call to the NRC Headquarter Operations Center.

Definitions:


HOSTILE ACTION: An act toward Salem or Hope Creek or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Salem or Hope Creek. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OWNER CONTROLLED AREA).



EAL#: HA1.1





EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HA1 Example EAL #1, #2 2. Salem – Hope Creek Security Contingency Plan 3. HC.OP-AB.SEC-0001 Security Event 4. HC.OP-AB.SEC-0002 Airborne Threat



EAL#: HA1.1



Contingency Event

Number



(UE ONLY) Yes / No






EAL#: HA1.1




EAL#: HS1.1



Initiating Condition: HOSTILE ACTION within the PROTECTED AREA


EAL# & Classification Level: HS1.1 – SITE AREA EMERGENCY

EAL:

A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Shift Manager or designee (Note 6)





Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment. Timely and accurate communications between Security Shift Manager and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan and Independent Spent Fuel Storage Installation Security Program. As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The SITE AREA EMERGENCY declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions. This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA1. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples



EAL#: HS1.1

include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the Salem – Hope Creek Security Contingency Plan (ref. 1).


The Security Shift Manager is defined as the Security Operations Supervisor or designee.

These individuals are the designated on-site personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Salem – Hope Creek Security Contingency Plan (Safeguards) information.

PROJECTILES that are directed into or that have impacted the PA from the OCA or beyond are considered under this EAL as HOSTILE ACTIONS within the PA.

Attached to this EAL Basis is a “Security Contingency Event Summary Table” that indicates which Security Contingency Events could result in Security Supervision determining that a HOSTILE ACTION is or has occurred and therefore classification at the ALERT or higher level should be made based on the location (OCA or PA) of the HOSTILE ACTION. Security events that do not involve a HOSTILE ACTION may result in Security Supervision determining that a SECURITY CONDITION exists and therefore an UNUSUAL EVENT classification should be made per EAL HU4.1.

Definitions:

HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.


HOSTILE ACTION: An act toward Salem or Hope Creek or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack



EAL#: HS1.1

on Salem or Hope Creek. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OCA).


PROJECTILE: An object that impacts Salem and/or Hope Creek that could cause concern for continued operability, reliability, or personnel safety.



EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HS1 Example EAL #1 2. Salem – Hope Creek Security Contingency Plan 3. HC.OP-AB.SEC-0001 Security Event 4. HC.OP-AB.SEC-0002 Airborne Threat



EAL#: HS1.1



Contingency Event

Number



(UE ONLY) Yes / No






EAL#: HU2.1


EAL Subcategory: 2 – Seismic Event

Initiating Condition: Seismic event greater than OBE levels



EAL:

Actuation of the OBE Seismic Switch (> 0.1 g) as indicated by EITHER:

Annunciator C6-C4 (SEISMIC MON PNL C673) activated

Amber alarm light on the Seismic Switch Power Supply Drawer Panel 10C673

Basis:

This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant. Event verification with external sources should not be necessary during or following an OBE. Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.1g). The Shift Manager or Emergency Coordinator may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.); however, the verification action must not preclude a timely emergency declaration. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.


Ground motion acceleration of 0.1g is the Operating Basis Earthquake (OBE) for HCGS.

HCGS seismic instrumentation consists of a Kinemetrics SMA-3 Strong Motion Accelerograph (Auxiliary Building Upper Relay Room) and associated sensors that are equipped with seismic triggers set to alarm and initiate recording at an acceleration equal to or exceeding 0.01 g.



EAL#: HU2.1

When the seismic trigger activates the SMA-3 Event Indicator (flag) will change from black to white and the amber event alarm will illuminate. The amber event alarm will extinguish when ground acceleration reduces below the 0.01 g setpoint but the Event Indicator (flag) will remain white until manually reset.

The amber Seismic Switch Event Alarm on the Seismic Switch Power Supply (SP-1) will illuminate at an acceleration equal to or exceeding 0.1 g (OBE). This also annunciates the seismic activity alarm C6-C4 (SEISMIC MON PNL C673). Three time-history triaxial acceleration sensors are provided. These sensors transmit electrical signals to the Upper Relay room where they are recorded on magnetic tape. These tapes are analyzed to determine the exact magnitude of the seismic event and to confirm whether the OBE has been exceeded.

The NEIC can be contacted by calling (303) 273-8500. Select option #1 and inform the analyst you wish to confirm recent seismic activity in the vicinity of Salem/Hope Creek Generating Station. Provide the analyst with the following coordinates: 39º 27' 46" (39.465°) north latitude, 75º 32' 08'' (75.537°) west longitude.

Alternatively go to the USGS NEIC website:

http://earthquake.usgs.gov/eqcenter/

On the US map, click on 'New Jersey' and then click on earthquake indicator for information. The maps are updated within 5 min. of a measured earthquake.

Additional Earthquake information can be found on the internet at:

http://www.earthquake.usgs.gov

http://www.mgs.md.gov (click on “Live Earthquake Data online”)

http://earthquake.usgs.gov/regional/neic

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HU2 Example EAL #1 2. HC.OP-SO.SG-0001(Z) Seismic Instrumentation System Operation 3. HC.OP-AB.MISC-0001(Q) Acts of Nature 4. HC.OP-AR.ZZ-0011(Q) Overhead Annunciator Window Box C6 5. UFSAR 2.1.1.1 Specification of Location 6. UFSAR 1.2.1.5 Seismology



EAL#: HU3.1


EAL Subcategory: 3 – Natural or Technological Hazard

Initiating Condition: Hazardous Event



EAL:

A tornado strike within the PROTECTED AREA

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. EAL HU3.1 addresses a tornado striking (touching down) within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.


A tornado touching down within the PROTECTED AREA warrants declaration of an UNUSUAL EVENT regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm.

Definitions:

PROTECTED AREA (PA): A security controlled area within the OWNER CONTROLLED AREA (OCA) that is enclosed by the security perimeter fence and monitored by intrusion detection systems. Access to the PA requires proper security clearance and is controlled at the Security Center.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HU3 Example EAL #1 2. HC.OP-AB.MISC-0001 (Q) Acts of Nature



EAL#: HU3.1




EAL#: HU3.2



Initiating Condition: Hazardous event


EAL# & Classification Level: HU3.2 – UNUSUAL EVENT

EAL:

Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current OPCON

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses FLOODING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.


Definitions:

FLOODING: A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary;



EAL#: HU3.2

(2) The capability to shut down the reactor and maintain it in a safe shutdown condition;

(3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HU3 Example EAL #2



EAL#: HU3.3






EAL:

Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill or toxic gas release)

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to IMPEDE the movement of personnel within the PROTECTED AREA. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.


As used here, the term "offsite" is meant to be areas external to the PROTECTED AREA.

Definitions:


PROTECTED AREA (PA): A security controlled area within the OWNER CONTROLLED AREA (OCA) that is enclosed by the security perimeter fence and monitored by intrusion detection systems. Access to the PA requires proper security clearance and is controlled at the Security Center.




EAL#: HU3.3




EAL#: HU3.4


EAL Subcategory: 3 – Natural & Destructive Phenomena




EAL:

A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Note 7: This EAL does NOT apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents.

Basis:

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant. This EAL addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles. Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc. This EAL is not intended to apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011. Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, S or C.




EAL#: HU3.4




EAL#: HU4.1


EAL Subcategory: 4 – Fire

Initiating Condition: FIRE potentially degrading the level of safety of the plant



EAL:

A FIRE is NOT extinguished within 15 min. of ANY of the following fire detection indications (Note 1):

Report from the field (i.e., visual observation) Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm

AND

FIRE is located in ANY Table H-1 area

Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded.

Table H-1 Fire Areas

Reactor Building

Control/Auxiliary Building

Service Water Intake Structure

Service/Radwaste Building

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. For EAL HU4.1 the intent of the 15-minute duration is to size the FIRE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.



EAL#: HU4.1

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication, or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial multiple alarms, indications, or report (including confirmation of a single fire alarm) was received, and not the time that a subsequent verification action was performed. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.


The Table H-1 Fire Areas include those plant structures identified as Seismic Category I.

Definitions:

FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HU4 Example EAL #1 2. UFSAR Table 3.2-1 HCGS Classification of Structures, Systems and Components



EAL#: HU4.2






EAL:

Receipt of a single fire alarm (i.e., NO other indications of a FIRE) AND

The fire alarm is indicating a FIRE within ANY Table H-1 area AND

The existence of a FIRE is NOT verified within 30 min. of alarm receipt (Note 1)


Table H-1 Fire Areas

Reactor Building

Control/Auxiliary Building

Service Water Intake Structure

Service/Radwaste Building

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e., proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed. A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to



EAL#: HU4.2

verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress. If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted. Basis-Related Requirements from Appendix R

Appendix R to 10 CFR 50, states in part:

Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."

When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.

Because fire may affect safe shutdown systems and because the loss of function of systems used to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in this EAL, the 30-minutes to verify a single alarm is well within this worst-case 1-hour time period. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.


The Table H-1 Fire Areas include those plant structures identified as Seismic Category I.

Definitions:


EAL Bases Reference(s):



EAL#: HU4.2

1. NEI 99-01, Rev. 06, HU4 Example EAL #2 2. UFSAR Table 3.2-1 HCGS Classification of Structures, Systems and Components



EAL#: HU4.2




EAL#: HU4.3






EAL:

A FIRE within the plant PROTECTED AREA NOT extinguished within 60 min. of the initial report, alarm or indication (Note 1)


Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.


Definitions:


PROTECTED AREA (PA): A security controlled area within the Owner-Controlled Area (OCA) that is enclosed by the security perimeter fence and monitored by intrusion detection systems. Access to the PA requires proper security clearance and is controlled at the Security Center.



EAL#: HU4.3

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HU4 Example EAL #3 2. UFSAR Table 3.2-1 HCGS Classification of Structures, Systems and Components



EAL#: HU4.4






EAL:

A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish

Basis:

This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant. If a FIRE within the plant PROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency (e.g., a local town Fire Department), then the level of plant safety is potentially degraded. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the FIRE is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions. Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA8.


Definitions:


PROTECTED AREA (PA): A security controlled area within the Owner-Controlled Area (OCA) that is enclosed by the security perimeter fence and monitored by intrusion detection systems. Access to the PA requires proper security clearance and is controlled at the Security Center.




EAL#: HU4.4

3. NEI 99-01, Rev. 06, HU4 Example EAL #4 4. UFSAR Table 3.2-1 HCGS Classification of Structures, Systems and Components



EAL#: HA5.1


EAL Subcategory: 5 – Hazardous Gases

Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown

OPCON Applicability: 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refueling


EAL:

Release of a toxic, corrosive, asphyxiant or flammable gas into ANY Table H-2 rooms or areas

AND

Entry into the room or area is prohibited or IMPEDED (Note 5)

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted

Table H-2 Safe Operation & Shutdown Rooms/Areas Bldg. – Ele. Rooms/Areas OPCON

Applicability


RB 77’ 4201 (10B242 MCC) 3, 4, 5



Basis:

This EAL addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown. This condition represents an actual or potential substantial degradation of the level of safety of the plant.



EAL#: HA5.1

An Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the gaseous release. The emergency classification is not contingent upon whether entry is actually necessary at the time of the release. Evaluation of the IC and EAL does not require atmospheric sampling; it only requires the Emergency Coordinator’s judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). An emergency declaration is not warranted if any of the following conditions apply: The plant is in an operating mode different than the mode specified for the affected

room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous release). For example, the plant is in Mode 1 when the gaseous release occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 3.

The gas release is a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., fire suppression system testing).

The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

If the equipment in the listed room or area was already inoperable, or out-of-service, before the event occurred, then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event.

An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment. This reduces the concentration of oxygen below the normal level of around 19%, which can lead to breathing difficulties, unconsciousness or even death. This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment.. Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.


The list of plant rooms or areas with entry-related mode applicability identified specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Rooms or



EAL#: HA5.1

areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 2).

This EAL should not be construed to include confined spaces that must be ventilated prior to entry or situations involving the fire department personnel who are using respiratory equipment during the performance of their duties unless it also affects personnel not involved with the fire department activates. In addition, those situations that require personnel to wear respiratory protection equipment as the result of airborne contamination as required by Radiation Protection personnel do not meet the intent of this EAL.

Definitions:


NOTE: EAL HA5.1 mode applicability has been limited to the applicable modes identified in Table H-2 Safe Operation & Shutdown Rooms/Areas. If due to plant operating procedure or plant configuration changes, the applicable plant modes specified in Table H-2 are changed, a corresponding change to Attachment 3 ‘Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases’ and to EAL HA5.1 mode applicability is required.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HA5 Example EAL #1 2. Attachment 7 Safe Operation & Shutdown Rooms/Areas Tables R-3 & H-2 Bases



EAL#: HA5.1




EAL#: HA6.1


EAL Subcategory: 6 – Control Room Evacuation

Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations



EAL:

An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel (RSP)

Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety. Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations. The necessity to control a plant shutdown from outside the Control Room, in addition to responding to the event that required the evacuation of the Control Room, will present challenges to plant operators and other on-shift personnel. Activation of the ERO and emergency response facilities will assist in responding to these challenges. Escalation of the emergency classification level would be via IC HS6.


Control Room evacuation represents a serious plant situation since the degree of plant control at the Remote Shutdown Panel (RSP) is not as complete as from the Control Room. The intent of this EAL is to declare an ALERT when the determination to evacuate the Control Room has been made based on environmental/personnel safety concerns, and the physical process of evacuating the Control Room per HC.OP-AB.HVAC-0002(Q), Control Room Environment, has commenced.

The Shift Manager (SM) determines if the Control Room requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions.



EAL#: HA6.1

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HA6 Example EAL #1 2. HC.OP-AB.HVAC-0002(Q) Control Room Environment 3. HC.OP-IO.ZZ-0008(Q) Shutdown from Outside the Control Room



EAL#: HS6.1


EAL Subcategory: 6 – Control Room Evacuation

Initiating Condition: Inability to control a key safety function from outside the Control Room

OPCON Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown, 4 – Cold Shutdown, 5 - Refueling


EAL:

An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel (RSP)

AND Control of ANY of the following key safety functions is NOT reestablished within 15 min. (Note 1):

Reactivity (modes 1 and 2 only) RPV water level RCS heat removal

Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded

Basis:

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time. The determination of whether or not “control” is established at the remote safe shutdown location(s) is based on Emergency Coordinator judgment. The Emergency Coordinator is expected to make a reasonable, informed judgment within 15 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s). Escalation of the emergency classification level would be via category F or IC CG1.



EAL#: HS6.1


The Shift Manager determines if the Control Room is inoperable and requires evacuation. Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threat in or adjacent to the Control Room, or other life threatening conditions.

The 15 minute clock for reestablishing control starts when the last licensed operator leaves the control room.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HS6 Example EAL #1 2. HC.OP-AB.HVAC-0002(Q) Control Room Environment 3. HC.OP-IO.ZZ-0008(Q) Shutdown from Outside the Control Room



EAL#: HU7.1


EAL Subcategory: 7 – EC Judgment

Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of an UNUSUAL EVENT



EAL:

Other conditions exist which, in the judgment of the Emergency Coordinator, indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an UNUSUAL EVENT.


Definitions:

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure: (1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result

in potential offsite exposures.



EAL#: HU7.1




EAL#: HA7.1



Initiating Condition: Other conditions exist that in the judgment of the Emergency Coordinator warrant declaration of an ALERT



EAL:

Other conditions exist which, in the judgment of the Emergency Coordinator, indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels

Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for an ALERT.


Definitions:

HOSTILE ACTION: An act toward Salem or Hope Creek or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate PSEG to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Salem or Hope Creek. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OCA).



EAL#: HA7.1

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HA7 Example EAL #1



EAL#: HS7.1



Initiating Condition: Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a SITE AREA EMERGENCY



EAL:

Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. ANY releases are NOT expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the site boundary.

Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for a SITE AREA EMERGENCY.


Definitions:




EAL#: HS7.1

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HS7 Example EAL #1



EAL#: HG7.1



Initiating Condition: Other conditions exist which in the judgment of the Emergency Coordinator warrant declaration of a GENERAL EMERGENCY


EAL# & Classification Level: HG7.1 – GENERAL EMERGENCY

EAL:

Other conditions exist which in the judgment of the Emergency Coordinator indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Coordinator to fall under the emergency classification level description for a GENERAL EMERGENCY.


Definitions:


IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.



EAL#: HG7.1

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, HG7 Example EAL #1



EAL#: SU1.1

EAL Category: S – System Malfunction

EAL Subcategory: 1 – Loss of AC Power

Initiating Condition: Loss of ALL offsite AC power capability to vital buses for 15 minutes or longer

OPCON Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown

EAL# & Classification Level: SU1.1 – UNUSUAL EVENT

EAL:

Loss of ALL offsite AC power capability to 4.16 KV vital buses for ≥ 15 min. (Note 1)


Basis:

This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC vital buses. This condition represents a potential reduction in the level of safety of the plant. For emergency classification purposes, “capability” means that an offsite AC power source(s) is available to the vital buses, whether or not the buses are powered from it. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Emergency Classification escalates to an ALERT under EAL SA1.1 based on AC power to 4.16 KV vital buses being reduced to a single source.


The AC power distribution is summarized in Attachment 2.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06 – SU1 Example EAL #1 2. 205415-A-8765 Sheet 1 500 kv Transmission Plan & Profiles 3. E-0001-0 sheet 1 Hope Creek Generating Station Single Line Diagram Station 4. UFSAR 8.1.2 Onsite Power Systems 5. HCGS Technical Specifications 3.8.1 AC Sources



EAL#: SU1.1

6. HC.OP-AB.ZZ-0135 (Q) Station Blackout / Loss of Offsite Power / Diesel Generator Malfunction



EAL#: SA1.1



Initiating Condition: Loss of ALL but one AC power source to vital buses for 15 minutes or longer


EAL# & Classification Level: SA1.1 – ALERT

EAL:

AC power capability to 4.16 KV vital buses reduced to a single power source for ≥ 15 min. (Note 1)

AND

ANY additional single power source failure will result in loss of ALL AC power to SAFETY SYSTEMS


Basis:

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. This IC provides an escalation path from IC SU1. An “AC power source” is a source recognized in AOPs and EOPs, and capable of supplying required power to a vital bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SS1.



EAL#: SA1.1


“Capability” means the power source can be aligned to provide power to a vital bus within 15 minutes or is currently supplying power to at least one vital bus.


This hot condition Alert EAL is equivalent to the cold condition Unusual Event EAL CU2.1.

Definitions:


(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could

result in potential offsite exposures.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06 – SA1 Example EAL #1 2. HC.OP-AB.ZZ-0135 (Q) Station Blackout / Loss of Offsite Power / Diesel Generator

Malfunction 3. 205415-A-8765 Sheet 1 500 kv Transmission Plan & Profiles 4. E-0001-0 sheet 1 Hope Creek Generating Station Single Line Diagram Station 5. UFSAR 8.1.2 Onsite Power Systems 6. HCGS Technical Specifications 3.8.1 A.C. Sources 7. HC.OP-EO.ZZ-0105 (Q) Reactor Pressure Vessel Control – HPCI/RCIC Only



EAL#: SS1.1



Initiating Condition: Loss of ALL offsite power and ALL onsite AC power to vital buses for 15 minutes or longer


EAL# & Classification Level: SS1.1 – SITE AREA EMERGENCY

EAL:

Loss of ALL offsite and ALL onsite AC power to 4.16 KV vital buses for ≥ 15 min. (Note 1)


Basis:

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC RG1, category F or IC SG1.


The intent of this EAL is to classify degraded AC power events that result in a loss of all offsite power sources to the 4.16 KV vital buses along with a loss of all onsite power sources (EDGs).



EAL#: SS1.1


This hot condition SITE AREA EMERGENCY EAL is equivalent to the cold condition ALERT EAL CA2.1.

Definitions:




EAL Bases Reference(s): 1. NEI 99-01, Rev. 06 – SS1 Example EAL #1 2. HC.OP-AB.ZZ-0135 (Q) Station Blackout / Loss of Offsite Power / Diesel Generator

Malfunction 3. 205415-A-8765 Sheet 1 500 kv Transmission Plan & Profiles 4. E-0001-0 sheet 1 Hope Creek Generating Station Single Line Diagram Station 5. UFSAR 8.1.2 Onsite Power Systems 6. HCGS Technical Specifications 3.8.1 AC Sources



EAL#: SG1.1



Initiating Condition: Prolonged loss of ALL offsite and ALL onsite AC power to vital buses


EAL# & Classification Level: SG1.1 – GENERAL EMERGENCY

EAL:

Loss of ALL offsite and ALL onsite AC power to 4.16 KV vital buses

AND EITHER:

Restoration of at least one vital 4.16 KV bus in < 4 hrs is NOT likely (Note 1)

RPV level CANNOT be restored and maintained > -185 in.


Basis:

This IC addresses a prolonged loss of all power sources to AC vital buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC vital bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers. The estimate for restoring at least one vital bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.



EAL#: SG1.1

The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.



Four hours is the station blackout coping time. Beyond the four hour window, RPV injection capability may no longer be available and degradation in core cooling will commence.

An extreme challenge to the ability to cool the core occurs when RPV water level cannot be maintained above -185”. Although this is below the Top of Active Fuel (Loss of Core Submergence), maintaining Reactor water level above -185” will ensure sufficient steam flow from the covered portion of the core to preclude fuel clad temperatures in the uncovered portion of the core from exceeding 1500°F. This is referred to as the Minimum Steam Cooling RPV Water Level (MSCRWL). Inability to maintain this level for an extended period of time could be precursors of a core melt sequence.

The RPV level threshold is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term “cannot be restored and maintained above” means the value of RPV water level is not able to be brought above the specified limit (-185 in.). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the limit, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the limit cannot be attained (ref 7, 8).

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06 – SG1 Example EAL #1 2. HC.OP-AB.ZZ-0135 (Q) Station Blackout / Loss of Offsite Power / Diesel Generator

Malfunction 3. 205415-A-8765 Sheet 1 500 kv Transmission Plan & Profiles 4. E-0001-0 sheet 1 Hope Creek Generating Station Single Line Diagram Station 5. UFSAR 8.1.2 Onsite Power Systems 6. HCGS Technical Specifications 3.8.1 AC Sources 7. HC.OP-EO.ZZ-0101(Q)-FC RPV Control 8. HC.OP-EO.ZZ-0101A(Q)-FC ATWS - RPV Control 9. HC.OP-EO.ZZ-0105 (Q) Reactor Pressure Vessel Control – HPCI/RCIC Only



EAL#: SS2.1


EAL Subcategory: 2 – Loss of DC Power

Initiating Condition: Loss of ALL vital DC power for 15 minutes or longer



EAL:

< 108 V DC bus voltage indication on ALL vital 125 V DC buses for ≥ 15 min. (Note 1)


Basis:

This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC RG1, category F or IC SG2.


Per Technical Specifications, 108 VDC is the minimum voltage required for operability of the Class 1E 125 VDC buses following a battery discharge test. Although continued equipment operation may occur with degraded voltage, this value signifies the minimum operable voltage allowed.

125 VDC Power Channels A-D provide control power to Engineered Safety Features actuation, diesel generator auxiliaries, plant alarm and indication circuits, as well as the control power for the associated loads. If 125 VDC power is lost for an extended period of time (greater than 15 minutes) critical plant functions such as 4.16 KV breaker controls, CS and RHR pump controls required to maintain safe plant conditions may not operate, and core uncovery with subsequent Reactor Coolant System (RCS) and Primary Containment failure might occur.

Loss of ADS may create a loss of low pressure ECCS availability due to the potential inability to depressurize the reactor. In addition, loss of these buses will eventually lead to MSIV closure and reactor scram due to the loss of the Primary Containment Instrument Gas (PCIG).



EAL#: SS2.1

Subsequent to MSIV closure, much of the equipment noted above will be required for plant stabilization and shutdown.

A sustained loss of 125 V DC power will threaten the ability to remove heat from the reactor core and from the primary containment. SRVs will remain operable in the relief mode and the heat addition to the primary containment could result in a loss of the Primary Containment as a fission product release barrier.

Loss of all vital 1E 125 V DC power will also render HPCI and RCIC inoperable for automatic initiation, and from the Control Room due to loss of control power.

Class 1E 125 VDC battery bank capacities are as follows:

Channel Switchgear Battery Capacity (design 4 hours)

A 10D410 1AD411 1885 AH at 8 hours

B 10D420 1BD411 1885 AH at 8 hours

C 10D430 1CD411 1885 AH at 8 hours

D 10D440 1DD411 1885 AH at 8 hours

C 10D436 1CD447 577 AH at 8 hours

D 10D446 1DD447 577 AH at 8 hours

In OPCONs 1, 2, or 3, the loss of any single channel 125V DC power source would require the channel to be restored within 2 hours or the unit placed in at least Hot Shutdown within the next 12 hours and in Cold Shutdown within the following 24 hours.

This Site Area Emergency EAL is the hot condition equivalent of the cold condition loss of DC power Unusual Event EAL CU5.1.

Definitions:






EAL#: SS2.1

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SS8 Example EAL #1 2. HC.OP-AB.ZZ-0150 (Q), 125VDC System Malfunction 3. HC.OP-AB.ZZ-0135 (Q), Station Blackout / Loss of Offsite Power / Diesel Generator

Malfunction 4. HCGS Technical Specifications Section 3/4.8.2.1, 3/4.8.3.1.b; DC Sources – Operating 5. UFSAR 8.3.2 DC Power Systems



EAL#: SS2.1




EAL#: SG2.1

EAL Category: S – System Malfunctions


Initiating Condition: Loss of ALL vital AC and vital DC power sources for 15 minutes or longer


EAL# & Classification Level: SG2.1 – GENERAL EMERGENCY

EAL:

Loss of ALL offsite and ALL onsite AC power to 4.16 KV vital buses for ≥ 15 min.

AND

< 108 V DC bus voltage indication on ALL vital 125 V DC buses for ≥ 15 min. (Note 1)


Basis:

This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both vital AC and vital DC power will lead to multiple challenges to fission product barriers. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.


The intent of this EAL is to classify degraded AC power events that result in a loss of all offsite power sources 13.8 KV to the 4.16 KV vital buses along with a loss of all onsite AC power sources (EDGs) concurrent with a loss of all 125 VDC vital power.




EAL#: SG2.1


Definitions:




EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SG8 Example EAL #1 2. HC.OP-AB.ZZ-0135 (Q) Station Blackout / Loss of Offsite Power / Diesel Generator

Malfunction 3. 205415-A-8765 Sheet 1 500 kv Transmission Plan & Profiles 4. E-0001-0 sheet 1 Hope Creek Generating Station Single Line Diagram Station 5. UFSAR 8.1.2 Onsite Power Systems 6. HCGS Technical Specifications 3.8.1 AC Sources 7. HC.OP-AB.ZZ-0150 (Q), 125VDC System Malfunction 8. HCGS Technical Specifications Section 3/4.8.2.1, 3/4.8.3.1.b; DC Sources – Operating 9. UFSAR 8.3.2 DC Power Systems



EAL#: SU3.1


EAL Subcategory: 3 – Loss of Control Room Indications

Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer



EAL:

An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for ≥ 15 min. (Note 1)


Table S-1 Safety System Parameters

Reactor power RPV water level RPV pressure Primary containment pressure Torus water level Torus temperature

Basis:

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL, an “inability to monitor” means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly



EAL#: SU3.1

impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via IC SA3.


Definitions:





EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SU2 Example EAL #1 2. HC.OP.AB.MISC-0002 (Q) CRIDS / Overhead Annunciators / PPC Computer 3. HC.OP-AR.ZZ-0011 Window C6-C5, SPDS System Trouble 4. HCGS Technical Specifications 3/4.3, Instrumentation



EAL#: SA3.1


EAL Subcategory: 3 – Loss of Control Room Indications

Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress



EAL:

An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for ≥ 15 min. (Note 1)

AND

ANY significant transient is in progress, Table S-2


Table S-1 Safety System Parameters

Reactor power RPV water level RPV pressure Primary containment pressure Torus water level Torus temperature

Table S-2 Significant Transients

Reactor scram Thermal power reduction > 25% Electrical load rejection > 22% ECCS injection Thermal power oscillations > 10%



EAL#: SA3.1

Basis:

This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an “inability to monitor” means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making. This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via category F or IC RS1


Significant transients are listed in Table S-2.

The Plant Process Computer System (PPC), Control Room Integrated Display System (CRIDS) and Safety Parameter Display System (SPDS) serve as redundant indicators which may be utilized as compensatory measures in lieu of the Control Room indicators associated with safety functions.



EAL#: SA3.1

Definitions:





EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SA2 Example EAL #1 2. UFSAR Section 1.2.4.3.6 Turbine Bypass System and Pressure Control System 3. HC.OP.AB.MISC-0002 (Q) CRIDS / Overhead Annunciators / PPC Computer 4. HC.OP-AR.ZZ-0011 Window C6-C5, SPDS System Trouble 5. HC.OP-AR.ZZ-0011 Window C6-B5, BOP Computer Trouble 6. HCGS Technical Specifications 3/4.3, Instrumentation



EAL#: SA3.1




EAL#: SU4.1


EAL Subcategory: 4 – RCS Activity

Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits



EAL:

Offgas Pretreatment Radiation Monitor (9RX621/9RX622) High alarm

Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via category F or the recognition category R ICs.


An Offgas Pretreatment Radiation Monitor High alarm is indicative of a degradation of the fuel clad, and is a precursor of a more serious problem. The alarm is set at 2.2E+04 mR/hr, which ensures that the alarm will actuate prior to exceeding the Technical Specification Offgas System Noble Gas Effluent Limit of 3.3E5 µCi/sec.

Classification should be based on an Offgas Pretreatment Radiation Monitor High Alarm actuating specifically due to fuel clad degradation, thus precluding unwarranted UNUSUAL EVENT declaration as the result of a resin / chemical intrusion transient, HWCI System malfunction, etc. UNUSUAL EVENT declaration is warranted only when actual fuel damage has occurred.

The Offgas Pretreatment Radiation Monitors (9RX621 / 9RX622) monitor gamma radiation levels attributable to the non-condensable fission product gases produced in the reactor and transported with steam through the turbine to the condenser. These instruments take a sample from the sample tap between the fourth and fifth holdup pipe of the Offgas system.

Restricting the gross radioactivity from the Main Condenser provides reasonable assurance that the whole body dose to an individual at the exclusion area boundary will not exceed a



EAL#: SU4.1

small fraction of the limits of 10 CFR 100 in the event this effluent is inadvertently discharged directly to the environment without treatment.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SU3 Example EAL #1 2. HC.OP-AR.ZZ-0011(Q) Overhead Annunciator Window Box C6 3. HCGS Technical Specification Table 3.3.7.1-1 Radiation Monitoring Instrumentation 4. HCGS Technical Specification 3.11.2.7 Radioactive Effluents - Main Condenser 5. HC.OP-AB.RPV-0008(Q) Reactor Coolant Activity 6. HC.RP-AR.SP-0001(Q) Radiation Monitoring System Alarm Response 7. OE-6144 Resin Intrusion, Hope Creek



EAL#: SU4.2


EAL Subcategory: 4 – RCS Activity

Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits



EAL:

Coolant activity > 4 µCi/gm dose equivalent I-131

Basis:

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.


A reactor coolant sample analysis with specific activity in excess of the Technical Specification limit of 4 µCi/gm Dose Equivalent Iodine-131 (DEI-131) is indicative of a degradation of the fuel clad, and is a precursor of more serious problems. This activity level is chosen instead of the 0.2 µCi/gm DEI-131 Technical Specification limit (under which operation is allowed to continue for up to 48 hours) to accommodate short duration Iodine spikes following changes in thermal power.

The Technical Specification limit on Reactor Coolant activity ensures that the 2 hour thyroid and whole body doses resulting from a Main Steam Line failure outside the primary containment during steady state operation will not exceed a small fraction of the 10 CFR 50.67 limits.

This limit accommodates Iodine spiking, which frequently occurs following shutdowns, startups, rapid power changes and coolant depressurization. Iodine spikes are characterized by a rapid increase in Reactor Coolant Iodine concentration by as much as three orders of magnitude followed by a return to pre-spike concentrations. This spiking is a temporary excursion and is not caused by a sudden fuel failure and should not be classified under this EAL provided coolant activity does not exceed the specified EAL threshold of 4 µCi/gm DEI- 131.



EAL#: SU4.2

Reactor Coolant Sample Activity of > 4 µCi/gm DEI- 131 can only occur as the result of fuel clad degradation and not as a result of Iodine spiking, resin / chemical intrusion transient, HWCI System malfunction, etc. UNUSUAL EVENT declaration is warranted only when actual fuel clad degradation has occurred.

The Technical Specification limit of > 100/E µCi/gm is excluded from this EAL because this limit does not include Iodine Activity.

Escalation to an ALERT or higher emergency classification occurs if a sample analysis of reactor coolant activity exceeds 300 µCi/gm DEI-131 via fission product barrier monitoring.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SU3, Example EAL #2 2. HGCS Technical Specifications 3.4.5 Specific Activity 3. HC.OP-AB.RPV-0008(Q) Reactor Coolant Activity



EAL#: SU5.1


EAL Subcategory: 5 – RCS Leakage

Initiating Condition: RCS Leakage for 15 minutes or longer



EAL:

RCS unidentified or pressure boundary leakage > 10 gpm for ≥ 15 min. OR

RCS identified leakage > 25 gpm for ≥ 15 min. OR

Leakage from the RCS to a location outside primary containment > 25 gpm for ≥ 15 min. (Note 1)

Note 1: The Emergency Coordinator should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded

Basis:

This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. The first and second EAL conditions are focused on a loss of mass from the RCS due to “unidentified leakage", "pressure boundary leakage" or "identified leakage” (as these leakage types are defined in the plant Technical Specifications). The third condition addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the primary containment or a location outside of primary containment. The leak rate values for each condition were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). The first condition uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage. The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification. A stuck-open Safety Relief Valve (SRV) or SRV



EAL#: SU5.1

leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL. The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible. Escalation of the emergency classification level would be via ICs of recognition category R or F.


Allowable leakage rates from the RCS are based on predicted and experimentally observed behavior of cracks in pipes. Utilizing the leak before break methodology, it is anticipated that there will be indication(s) of minor RCS boundary leakage prior to a fault escalating to a major leak or a system rupture. Detection of low levels of leakage while pressurized allows for implementation of mitigative actions and permits monitoring for catastrophic failure or rupture precursors.

The limit for RCS Unidentified Leakage and Pressure Boundary Leakage is set to a lower value than Identified Leakage due to concern over break propagation resulting from a small break that could potentially lead to a significantly larger loss of inventory.

RCS leakage is detected by monitoring:

Drywell atmospheric gaseous radioactivity

Drywell floor and equipment drain sump flow rate

Drywell air cooler condensate flow rate

Drywell pressure

RPV head flange leak detection system

Drywell temperature

Drywell Leak Detection (DLD) instrumentation is available via the Radiation Monitoring System (RM-11) to help determine Identified Leakage. Redundant instrumentation for DLD is available on Panel 10C604 located in the back of the Control Room.



EAL#: SU5.1

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SU4 Example EAL #1, #2, #3 2. HCGS Technical Specifications, Definitions 3. HCGS Technical Specifications 3.4.3.2 Operational Leakage 4. HC.OP-AB.CONT-0001 (Q) Drywell Pressure 5. HC.OP-AB.CONT-0002 (Q) Primary Containment 6. HC.OP-EO.ZZ-0101(Q)-FC Reactor/Pressure Vessel (RPV) Control 7. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 8. HC.OP-GP.ZZ-0005(Q) Drywell Leakage Source Detection 9. HC.OP-SO.SM-0001(Q) Isolation Systems Operation 10. UFSAR 5.2.5 Detection of Leakage Through the Reactor Coolant Pressure Boundary and

from the ECCS 11. UFSAR 7.6.1.3 Leak Detection System - Instrumentation and Control 12. HC.OP-AB.CONT-0006(Q) Drywell Leakage



EAL#: SU5.1




EAL#: SU6.1


EAL Subcategory: 6 – RPS Failure

Initiating Condition: Automatic or manual scram fails to shut down the reactor

OPCON Applicability: 1 - Power Operation, 2 - Startup


EAL:

An automatic or manual scram did NOT shut down the reactor as indicated by reactor power > 4% after ANY RPS setpoint is exceeded or manual scram action was initiated

AND A subsequent automatic scram or manual scram action taken at the reactor control console (mode switch, manual scram pushbuttons, manual ARI actuation) is successful in shutting down the reactor as indicated by reactor power ≤ 4% (Note 8)

Note 8: A manual scram action is ANY operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does NOT include manually driving in control rods or implementation of boron injection strategies

Basis:

This EAL addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant’s decay heat removal systems. If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram [using a different switch]. Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant’s decay heat removal systems.



EAL#: SU6.1

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be “at the reactor control consoles”. Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6. Depending upon the plant response, escalation is also possible via category F. Absent the plant conditions needed to meet either IC SA6 or category F, an Unusual Event declaration is appropriate for this event. Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and should be evaluated.

If the signal does not cause a plant transient and the scram failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.


The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is “fail safe”, that is it de-energizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical/pneumatic failure) or a failure of the Control Rod Drive System to permit the control rods to insert (hydraulic failure).

The Alternate Rod Insertion (ARI) function of the Redundant Reactivity Control System (RRCS) provides an automatic backup function for an electrical/pneumatic failure of RPS. A successful scram due to ARI following a failure of RPS would still be classified under this EAL because of the potentially serious consequences of an RPS failure.

The APRM downscale trip setpoint is 4%. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. At or below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown.



EAL#: SU6.1

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SU5 Example EAL #1 & #2 2. Technical Specifications 1.0 Definitions 3. Technical Specifications SL/LSSS 2.1/2.2 Safety Limits/ Limited Safety System Settings 4. Technical Specifications LCO 3/4.1 Reactor Control Systems 5. Technical Specifications LCO 3/4.3 Instrumentation 6. HC.OP-AB.IC-0003(Q) Reactor Protection System 7. HC.OP-EO.ZZ-0101A(Q)-FC ATWS – RPV Control



EAL#: SU6.1




EAL#: SA6.1



Initiating Condition: Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are NOT successful in shutting down the reactor

OPCON Applicability: 1 - Power Operation, 2 - Startup


EAL:

An automatic or manual scram fails to shut down the reactor as indicated by reactor power > 4%

AND

Manual scram actions taken at the reactor control console (mode switch, manual scram pushbuttons, manual ARI actuation) are NOT successful in shutting down the reactor as indicated by reactor power > 4% (Note 8)

Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does NOT include manually driving in control rods or implementation of boron injection strategies

Basis:

This EAL addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS. A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at backpanels or other locations within the Control Room, or any location outside the Control Room, are not considered to be “at the reactor control consoles”. Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.



EAL#: SA6.1

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS6. Depending upon plant responses and symptoms, escalation is also possible via category F. Absent the plant conditions needed to meet either IC SS6 or category F, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the recognition category F; however, this IC and EAL are included to ensure a timely emergency declaration.




EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SA5 Example EAL #1 2. Technical Specifications 1.0 Definitions 3. Technical Specifications SL/LSSS 2.1/2.2 Safety Limits/ Limited Safety System Settings 4. Technical Specifications LCO 3/4.1 Reactor Control Systems 5. Technical Specifications LCO 3/4.3 Instrumentation 6. HC.OP-AB.IC-0003(Q) Reactor Protection System 7. HC.OP-EO.ZZ-0101A(Q)-FC ATWS – RPV Control



EAL#: SS6.1



Initiating Condition: Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal

OPCON Applicability: 1 - Power Operations, 2 - Startup


EAL:

An automatic or manual scram fails to shut down the reactor as indicated by reactor power > 4%

AND

ALL actions to shut down the reactor are NOT successful as indicated by reactor power > 4%

AND EITHER:

RPV level CANNOT be restored and maintained > -185 in.

HCTL exceeded (EOP Curve SPT-P)

Basis:

This EAL addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency. In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the recognition category F ICs/EALs. This is appropriate in that the recognition category F ICs/EALs do not address the additional threat posed by a failure to shut down the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. Escalation of the emergency classification level would be via IC RG1 or category F.




EAL#: SS6.1



Reactor shutdown achieved by use of the alternate control rod insertion methods of EOP-101A Step RC/Q-20 is also credited as a successful manual action provided reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist.

An extreme challenge to the ability to cool the core occurs when RPV water level cannot be maintained above -185”. Although this is below the Top of Active Fuel (Loss of Core Submergence), maintaining Reactor water level above -185” will ensure sufficient steam flow from the covered portion of the core to preclude fuel clad temperatures in the uncovered portion of the core from exceeding 1500°F. This is referred to as the Minimum Steam Cooling RPV Water Level (MSCRWL). Inability to maintain this level for an extended period of time could be precursors of a core melt sequence.

The term “cannot be restored and maintained above” means the value of RPV water level is not able to be brought above the specified limit (-185 in.). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the limit, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the limit cannot be attained.

An extreme challenge to the primary containment occurs when heat cannot be removed from the primary containment resulting in elevated suppression pool water temperature. The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool water temperature from which a depressurization will not raise suppression chamber pressure above the Primary Containment Pressure Limit (PCPL) of 65 psig before the rate of energy transfer from the RPV to the primary containment is within the capacity of the primary containment vent. When the PCPL is challenged, primary containment venting will be required even if offsite radioactivity release rate limits will be exceeded.

The HCTL is a function of RPV pressure and suppression pool water temperature and is a measure of the maximum heat load, which the primary containment can withstand. Plant parameters in excess of the HCTL could be precursor of primary containment failure. The



EAL#: SS6.1

Heat Capacity Temperature Limit is given in Curve SPT-P of HC.OP-EO.ZZ-0102(Q), Primary Containment Control.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SG5 Example EAL #1 2. Technical Specifications 1.0 Definitions 3. Technical Specifications SL/LSSS 2.1/2.2 Safety Limits/ Limited Safety System Settings 4. Technical Specifications LCO 3/4.1 Reactor Control Systems 5. Technical Specifications LCO 3/4.3 Instrumentation 6. HC.OP-AB.IC-0003(Q) Reactor Protection System 7. HC.OP-EO.ZZ-0101A(Q)-FC ATWS – RPV Control 8. HC.OP-EO.ZZ-0206A(Q)-FC ATWS - RPV Flooding 9. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control



EAL#: SS6.1




EAL#: SU7.1


EAL Subcategory: 7 – Loss of Communications

Initiating Condition: Loss of ALL onsite or offsite communications capabilities



EAL:

Loss of ALL Table S-3 onsite communication methods OR

Loss of ALL Table S-3 offsite communication methods OR

Loss of ALL Table S-3 NRC communication methods

Table S-3 Communication Methods

System Onsite Offsite NRC

Direct Inward Dial System (DID) X X X

Station Page System (Gaitronics) X

Station Radio System X

Nuclear Emergency Telephone System (NETS) X X

Centrex Phone System (ESSX) X X

NRC (ENS) X



EAL#: SU7.1

Basis:

This EAL addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the States of New Jersey and Delaware. The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.


Onsite, offsite and NRC communications include one or more of the systems listed in Table S-3.

Direct Inward Dial System (DID)

Direct Inward Dial (DID) system is named for the dominant feature of the commercial telephone service provided by the local telephone company for the site. DID allows station telephones to be extensions or tied lines of the same systems. These exchanges can take advantage of backup power supplies provided to the stations, and may use either PSEG microwave, commercial telephone system microwave, or buried cable transmission systems to maintain external communications. This commercial telephone service is available as an additional backup for the NETS and Centrex/ESSX 1 system.

Station Page System (Gaitronics)

Gaitronics is a completely transistorized voice communication system with five voice channels: one page and five party. The system is designed for use in extreme environmental conditions such as dust, moisture, heat and noise. The system consists of handsets, speakers and their associated amplifiers. The power for this system is 120 volts AC from an inverted DC source to provide reliable communications during an emergency.



EAL#: SU7.1

Station Radio System

The Operations and Fire Protection Department UHF radio system is a multi-frequency system used routinely by both station Operations Departments and the Fire Protection Department. When an emergency event is declared, these radio frequencies serve both station Operations Support Centers (OSC).

Nuclear Emergency Telephone System (NETS)

The Nuclear Emergency Telecommunications System (NETS) is a privately controlled, self-contained telephone exchange that operates as a closed system, not accessible from other phone exchanges. This feature allows the system to be dedicated to emergency response use. The system may use PSEG microwave, commercial telephone system microwave, fiber optics, or buried cable transmission as needed. The exchange switching equipment is maintained at the Environmental & Energy Resource Center (EERC). As an independent system with an uninterruptible power supply, it may operate with or without local phone service or external power.

Centrex Phone System (ESSX)

The Centrex/Electronic Switch System Exchange 1 (Centrex/ESSX 1) is also a privately controlled exchange, which PSEG operates with its own microwave signal system. This system is also independent of local phone service, since each circuit is independently wired. The microwave signal is generated from corporate facilities in Newark, NJ, separated from any local effects of weather or telephone use. The exchange is accessible from other exchanges, but circuits are located only in PSEG facilities. It is considered the primary backup for the NETS system.

NRC (ENS)

The Emergency Notification System (ENS) is a dedicated communications system with the NRC, which is part of the Federal Telecommunications System (FTS) and consists of direct lines to the NRC. FTS lines are used to provide general accident information. These telephones are installed in the Control Room, TSC, and the EOF.

This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SU6 Example EAL #1, #2, #3 2. PSEG Nuclear Emergency Plan, Section 7 Communications 3. UFSAR 9.5.2 Communications Systems



EAL#: SU7.1




EAL#: SA8.1


EAL Subcategory: 8 – Hazardous Event Affecting Safety Systems

Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode



EAL:

The occurrence of ANY Table S-4 hazardous event AND EITHER: Event damage has caused indications of degraded performance in at least one train

of a SAFETY SYSTEM needed for the current operating mode The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or

structure needed for the current operating mode

Table S-4 Hazardous Events

Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics

as determined by the Shift Manager



EAL#: SA8.1

Basis: Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not visible damage and does not meet the intent of this EAL. This EAL is intended to consider the collateral damage that has occurred due to the equipment failure. This EAL addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. The first condition addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. The second condition addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via category F or IC RS1.


Definitions:

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.


FLOODING: A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:



EAL#: SA8.1


result in potential offsite exposures. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not visible damage.

This EAL is the hot condition equivalent of the cold condition EAL CA6.1.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, SA9 Example EAL #1 2. HC.OP-AB.MISC-001(Q) Acts of Nature 3. HC.FP-PS.KC-0000 Fire and Medical Emergency Response Manual



EAL#: SA8.1




EAL#: FB1.P

EAL Category: F – Fission Product Barrier Degradation

Subcategory: RPV Water Level

Initiating Condition: Potential Loss of Fuel Clad


EAL# & Point Value: FB1.P (4 points)

EAL:

RPV level CANNOT be restored and maintained above -161 in.

OR

RPV level CANNOT be determined

Basis:

This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. The RPV water level threshold is the same as RCS barrier Loss threshold RB1.L. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a SITE AREA EMERGENCY. This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory. The term “cannot be restored and maintained above” means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and



EAL#: FB1.P

maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA6 or SS6 will dictate the need for emergency classification. Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified.


Core Submergence is the preferred method of maintaining adequate core cooling. When RPV level decreases to below TAF, the ability to effectively remove decay heat is being challenged, and as such the Fuel Clad fission product barrier can no longer be considered intact. While the Emergency Operating Procedures provide contingencies to establish adequate core cooling when RPV level drops below TAF (Core Spray or Steam Cooling with or without injection), these actions are designed to be an alternative method of providing adequate core cooling while actions are taken to reestablish core submergence. Sustained partial or total core uncovery can result in fuel damage and a significant release of fission products to the Reactor coolant. Sustained core uncovery can also result in a breach of the RPV due to core melt material interaction with the RPV.

A Loss of Core Submergence will occur when the rate of inventory loss is greater than the rate of inventory makeup from high pressure injection sources. This condition can occur as the result of the following events/sequences:

• A LOCA will cause RPV level to reach the Top of Active Fuel when the LOCA is the result of a large break (momentary core uncovery is expected to occur under this condition) or when the LOCA is due to a small or intermediate break in combination with an inability of high pressure injection sources to keep up with the leakrate.

• A loss of high pressure injection sources without the presence of a LOCA will also result in RPV level decreasing to TAF, due to continued Reactor Steam Flow without makeup.

Either of these events/sequences results in a challenge to the Fuel Clad Barrier when RPV level CANNOT be restored AND maintained above the TAF due to core uncovery, hence classification at this threshold is appropriate. However, for both these sequences, Low Pressure ECCS are designed to inject to the RPV as RPV Pressure decreases below the shutoff head of the pumps. RPV Depressurization will occur either due to the LOCA or Manual initiation of Emergency RPV Depressurization when RPV level cannot be restored and



EAL#: FB1.P

maintained above -185 in., provided injection systems are available. This will allow for restoration of RPV level and re-establishment of Core Submergence.

Until the wide range RPV level (WR) indicator is downscale or becomes questionable, it should be used to assure RPV level is above TAF. Once WR is downscale, use compensated fuel zone indication (SPDS or graph) to assess RPV level above TAF and when compensated fuel zone level indication goes below TAF or becomes questionable, declare the EAL.

If RPV level cannot be determined, the indicated level value and trend have become unreliable to the extent that decisions concerning adequate core cooling cannot be made. RPV Flooding in accordance with EOP 206 or EOP 206A is required to assure continued core cooling. These EOPs specify actions to rapidly depressurize the RPV and establish sufficient RPV injection to cool the core by either steam cooling (with injection) or flooding the RPV to the main steam line penetrations. Entry to the RPV Flooding EOPs warrants classification of a SITE AREA EMERGENCY based on EAL RB1.L and this EAL.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Fuel Clad Potential Loss 2.A 2. Technical Specifications Table 3.3.3-2 ECCS Actuation Instrumentation Setpoints 3. Technical Specifications Figure B3/4-3-1 Reactor Vessel Water Level 4. HC.OP-EO.ZZ-0101(Q)-FC RPV Control 5. HC.OP-EO.ZZ-0206(Q)-FC RPV Flooding 6. HC.OP-EO.ZZ-0206A(Q)-FC ATWS - RPV Flooding 7. HC.OP-EO.ZZ-0101A(Q)-FC ATWS - RPV Control



EAL#: FB1.P

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EAL#: FB2.P


Subcategory: EC Judgment

Initiating Condition: Potential Loss of Fuel Clad


EAL# & Point Value: FB2.P (4 points)

EAL:

ANY condition in the opinion of the Emergency Coordinator that indicates potential loss of the Fuel Clad barrier

Basis:

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.


The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include IMMINENT barrier degradation, barrier monitoring capability and dominant accident sequences.

• Barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

• Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.



EAL#: FB1.L



Initiating Condition: Loss of Fuel Clad


EAL# & Point Value: FB1.L (5 points)

EAL:

SAG entry is required

Basis:

The Loss threshold represents any EOP requirement for entry into the Severe Accident Guidelines. This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured.


Core submergence is the mechanism of core cooling whereby each fuel element is completely covered with water. Indicated RPV level at or above the top of the active fuel (TAF) constitutes the principal means of confirming the adequacy of core cooling via this mechanism. Assurance of continued adequate core cooling through core submergence is achieved when RPV level can be maintained at or above TAF.

Spray cooling is the mechanism of core cooling whereby the uncovered portion of the core is cooled by spray flow. Adequate spray cooling exists by design when at least one Core Spray loop is operating at design flow (6150 gpm) and RPV level is at or above the elevation of the jet pump suctions (-215 in.). The covered portion of the core is then cooled by submergence while the uncovered portion is cooled by spray flow.

Steam cooling is the mechanism of core cooling whereby steam updraft up through the uncovered portion of each fuel bundle is sufficient to prevent the temperature of the hottest fuel rod from exceeding the appropriate limiting value, which is specific to the mode of steam cooling being employed (i.e., with and without injection of makeup water to the RPV).

With injection of makeup water into the RPV established, adequate core cooling exists when steam flow through the core is sufficient to preclude the peak clad temperature of the hottest fuel rod from exceeding 1500°F, the threshold temperature for fuel rod perforation. Assurance of continued adequate core cooling is achieved when RPV level can be maintained at or above the Minimum Steam Cooling RPV Water Level (-185 in.).



EAL#: FB1.L

With no injection into the RPV established, adequate core cooling exists only so long as the covered portion of the reactor core generates sufficient steam to preclude the peak clad temperature of the hottest fuel rod from exceeding 1800°F, the threshold temperature for significant metal-water reaction.

Prolonged lack of cooling may result in severe overheating of the fuel clad, additional release of energy from accelerated clad oxidation, and eventual fuel melting. For events starting from full power operation, the failure to promptly reflood could result in some fuel melting. Even under these conditions vessel failure and primary containment failure with resultant release to the public would not be expected for some time. Reactor Water Level remaining below TAF for an extended amount of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions.

Ample time should be allowed for Low Pressure ECCS and alternate injection systems to restore RPV level prior to entry into this classification. The time basis for deciding whether or not RPV level can be maintained > -185 in. should be based on the rate of reactor depressurization, the availability of low-pressure injection sources, (ECCS and alternate injection systems), and the rate of Reactor coolant inventory loss. Indications such as RPV level trend, injection flow rates, primary containment parameter trends, and low pressure injection system capability should also be considered.

In the event, RPV level cannot be restored > -185 in., adequate core spray flow cannot be established or RPV water level cannot be determined with indication of fuel damage, SAG entry will be required by the EOPs.

This threshold is also a Potential Loss of the Containment barrier (CB1.L). Since SAG entry occurs after core uncovery has occurred, a Loss of the RCS barrier exists (RB1.L). SAG entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a GENERAL EMERGENCY classification.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Fuel Clad Loss 2.A 2. HC.OP-EO.ZZ-0101A(Q)-FC ATWS – RPV Control 3. HC.OP-EO.ZZ-0101(Q)-FC RPV Control 4. HC.OP-EO.ZZ-0206(Q)-FC RPV Flooding 5. HC.OP-EO.ZZ-0206A(Q)-FC ATWS - RPV Flooding 6. HC.OP-AM.ZZ-0001(Z) Severe Accident Guidelines (SAG) 7. EP FAQ 2015-001



EAL#: FB2.L


Subcategory: PC Radiation / RCS Activity




EAL:

DAPA radiation monitor RI-4825A or RI-4825B reading EITHER of the following:

• With drywell sprays, ≥ 2,000 R/hr

• Without drywell sprays, ≥ 4,000 R/hr

Basis:

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.


Drywell Atmosphere Post Accident (DAPA) Radiation monitors indicating 2000 R/hr or greater corresponds to a release of Reactor Coolant to the drywell with a concentration of 300 µCi/gm Dose Equivalent Iodine-131 (DEI-131) into the Primary Containment when Drywell Spray is in service. This value of reactor coolant activity is well above the threshold that could occur as the result of Iodine spiking, resin/chemical intrusion transients or a HWCI System malfunction. This activity level corresponds to fuel damage of approximately 4%.

EAL FB3.L provides a core damage analysis showing that a Reactor Coolant activity of 300 µCi/gm Dose Equivalent Iodine-131(DEI) is indicative of 4% fuel damage. Using Attachment 2 of HC.EP-EP.ZZ-0205, 4% fuel damage is indicated by a DAPA reading of approximately 2000 R/hr at 1 hour after shutdown (the most conservative) with drywell spray in service and approximately 4000 R/hr at 1 hour after shutdown with drywell spray not in service.

From Attachment 2 of HC.EP-EP.ZZ-0205, Step 6 the following formula was used to derive the EAL set-points:



EAL#: FB2.L

% Cladding Damage= �Indicated Radiation Level

100% Cladding Damage Radiation Level�×100

% 4 (% Cladding Damage) = DAPA Rad Monitor Reading X100 50,000R/HR (from Att 2. Fig. 1, of HC.EP-EP.ZZ-0205 with Sprays) DAPA Reading = (4) (50,000R/HR ) 100

DAPA Reading = 2000R/hr (with Drywell Spray in service)

OR

4 (% Cladding Damage) = DAPA Rad Monitor Reading X100 100,000R/HR (from Att 2. Fig. 1, of HC.EP-EP.ZZ-0205 without Sprays)

DAPA Reading = (4) (100,000R/HR ) 100

DAPA Reading = 4000R/hr (without Drywell Spray in Service)

DAPA radiation monitors RE-4825A and RE-4825B are high range area radiation monitors and are located in the drywell at 145’ elevation. Each detector has a range of 1 to 108 R/hr and a built in check source that keeps the associated channel on-scale at approximately 1.2 R/hr. In the Control Room, DAPA radiation levels are displayed on RI-4825A (RM-11 9RX635) and B (RM-11 9RX636) (Panel 10C604) and RR-4825A and B (10C650). Drywell radiation levels ≥ 2.0E+2 R/hr activates Overhead Annunciator C6-C1, RADIATION MONITORING ALARM/TRBL.

A structural I-beam in the drywell shields DAPA radiation monitor RE-4825B from shine due to radioactivity in the RCS piping. RE-4825A is not shielded. When the RCS is intact, elevated RCS coolant activity produces higher readings on RE-4825A than on RE-4825B. During events in which radioactivity is dispersed into the drywell atmosphere, structural shielding diminishes as the monitors become immersed in the radiation field. If the difference in RE-4825A and RE-4825B readings should decrease, it may be inferred that a LOCA event is occurring.

In order to facilitate the EAL declaration, the EAL threshold that specifies, “With Drywell Sprays, ≥ 2000R/Hr” should be used if Drywell Spray has been or is in service at any time during the event. It can be assumed that Drywell Spray in service for any amount of time results in some lowering of Drywell Radiation and therefore use of the lower threshold value is the correct EAL threshold to use.



EAL#: FB2.L

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Fuel Clad Loss 4.A 2. HC.EP-EP.ZZ-0205(Q) TSC – Post Accident Core Damage Assessment 3. UFSAR Table 11.5-1 Hope Creek Radiation Monitoring System 4. HC.OP-AR.ZZ-0011(Q) Overhead Annunciator Box C6



EAL#: FB2.L




EAL#: FB3.L


Subcategory: PC Radiation / RCS Activity




EAL:

Primary coolant activity > 300 µCi/gm dose equivalent I-131

Basis:

This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.


The percentage of Fuel Damage that corresponds to an RCS Activity of 300 µCi/gm DEI-131 is calculated as follows (for purposes of this calculation, cc and gm are considered equivalent):

Dose Factors (RG-1.109)

I-131 = 4.39E-3 I-132 = 5.23E-5 I-133 = 1.04E-3 I-134 = 1.37E-5 I-135 = 2.14E-4

Total core inventory (HCGS-UFSAR, table 12.2-135). This table gives 50% inventory, so table values are multiplied by 2.0.

I-131 = 8.64E7 Ci I-132 = 1.29E8 Ci



EAL#: FB3.L

I-133 = 1.99E8 Ci I-134 = 2.32E8 Ci I-135 = 1.81E8 Ci

Reactor Water Volume = 13000 cubic feet (HCGS Calculation BJ-0024)

Clad Release Fraction for iodines = 0.02 (Table 4.1, NUREG-1228)

The activity of each isotope in the clad would then be:

I-131 = 8.64E7(0.02) = 1.73E6 Ci I-132 = 1.29E8(0.02) = 2.58E6 Ci I-133 = 1.99E8(0.02) = 3.98E6 Ci I-134 = 2.32E8(0.02) = 4.64E6 Ci I-135 = 1.81E8(0.02) = 3.62E6 Ci

These activities are equivalent to 2.89E6 Ci DEI-131

)339.4( 4)(3.62E6)-(2.14E + 5)(4.64E6)-(1.37E + 3)(3.98E6)-(1.04E + 5)(2.58E6)-(5.23E + 3)(1.73E6)-(4.39E131

-=-

EDEI

Calculating the equivalent concentration:

which represents the 100% fuel damage concentration.

300 µCi/cc DEI-131 is then equivalent to:

This is rounded to 3.8%.

Conc = 2.89E6 Ci(1E6 Ci / Ci) 13000 cf(2.8E4 cc / cf)

µ = 7.94E3 µCi/cc

300 Ci / cc

7.94E3 Ci / cc µµ = 3.78%



EAL#: FB3.L

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Fuel Clad Loss 1.A 2. HC.OP-AB.RPV-0008 (Q) Reactor Coolant Activity 3. HCGS Technical Specification LCO 3.4.5 Specific Activity 4. NUREG 1228 - Source Term Estimation During Incident Response to Severe Nuclear

Power Plant Accidents, Table 4.1 5. Reg. Guide 1.109, Table E-9 6. UFSAR Table 12.2-135 Post Accident Source Terms for Depressurized Liquid Containing

Systems



EAL#: FB3.L




EAL#: FB4.L

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Fuel Clad Loss 6.A



EAL#: RB1.P


Subcategory: RCS Leak Rate

Initiating Condition: Potential Loss of RCS


EAL# & Point Value: RB1.P (4 points)

EAL:

UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

ANY EOP 103 Reactor Bldg room temperature Table 1, Column 1

ANY EOP 103 Reactor Bldg local rad monitoring alarm

EOP 103 TABLE 1

Area Description & Room NumberColumn 1

Max NormalOp Temp

CRD Pump Room (4202) 115°F 140°F

HPCI (4111)

Core Spray Pump Rooms A(4118) & C(4116)

RHR Pump Rooms A(4113) & C(4114)

SACS A & C (4309)

RCIC Pump Room (4110)

Core Spray Pump Rooms B(4104) & D(4105)

RHR Pump Rooms B(4109) & D(4107)

SACS B & D (4307)

RWCU Pipe Chase (4402)

115°F 250°F

115°F 140°F

115°F 140°F

115°F 140°F

115°F 250°F

115°F 140°F

115°F 140°F

115°F 140°F

160°F 160°F

Column 2Max Safe Op Temp

Basis:

Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, HPCI, etc., which indicate a direct path from the RCS to areas outside primary containment.



EAL#: RB1.P

A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an ALERT classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a SITE AREA EMERGENCY when combined with Containment Barrier Loss threshold CB5.L (after a containment isolation) and a GENERAL EMERGENCY when the Fuel Clad Barrier criteria is also exceeded.


The Reactor Building room temperatures listed in EOP 103 Table 1, Column 1, and the Reactor Building local rad monitoring alarm levels define the Maximum Normal Operating values because they signify the onset of abnormal system operation. When parameters reach these levels, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP 103.

In general, multiple indications should be used to determine if a primary system is discharging outside primary containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.

This threshold allows for valve closure to isolate any systems not completely isolated, prior to event classification. Isolation is defined as the closure of any valve in the system(s) not completely isolated. For example, if the isolation logic fails to cause valve closure, but operator actions successfully isolates the containment breach/leak path, then classification under this threshold is not warranted. This includes manual and motor operated valves either locally or remotely.

Although this EAL allows for valve closure both remotely and locally, the time to attempt closure and make a decision if leak isolation was successful runs concurrently with the EAL 15-minute assessment clock:

If, during the EAL 15-minute assessment period attempts to isolate the leak are successful then, this threshold is NOT exceeded and classification per this threshold should NOT be made.



EAL#: RB1.P

If, during the EAL 15-minute assessment period attempts to isolate the leak are NOT

successful then, this threshold is exceeded and classification should be made at that time.

If near the end of the 15 minute assessment period and the operators have not been

able to attempt leak isolation or the EC is not convinced that an isolation attempt has been successful, then this threshold is exceeded and classification should be made at or before the 15-minute assessment time expires.

Definitions:

UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 RCS Barrier Potential Loss 3.A 2. HC.OP-EO.ZZ-0103/4(Q)-FC Reactor Building and Rad Release Control



EAL#: RB1.P




EAL#: RB2.P



Initiating Condition: Potential Loss of RCS


EAL# & Point Value: RB2.P (4 points)

EAL:

ANY condition in the opinion of the Emergency Coordinator that indicates potential loss of the RCS barrier

Basis:

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the RCS Barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.


The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include IMMINENT barrier degradation, barrier monitoring capability and dominant accident sequences.

Barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.



EAL#: RB2.P

Definitions:


EAL Bases Reference(s): 1. NEI 99-01, Rev. 05, Table 9-F-2 RCS Potential Loss 6.A



EAL#: RB1.L



Initiating Condition: Loss of RCS


EAL# & Point Value: RB1.L (5 points)

EAL:

RPV level CANNOT be restored and maintained above -161 in.

OR

RPV level CANNOT be determined

Basis:

This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.

The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold FB1.L. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a SITE AREA EMERGENCY.

This threshold is considered to be exceeded when, as specified in the EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

The term, “cannot be restored and maintained above,” means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and



EAL#: RB1.L

maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs SA5 or SS5 will dictate the need for emergency classification.


The top of the active fuel is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and Containment (PC) barriers, and initiation of all ECCS. If RPV level cannot be maintained above the top of active fuel, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend.

Core Submergence is the preferred method of maintaining adequate core cooling. When RPV Level decreases to below TAF, the ability to effectively remove decay heat is being challenged, and as such the Fuel Clad fission product barrier can no longer be considered intact. While the Emergency Operating Procedures provide contingencies to establish adequate core cooling when RPV Level drops below TAF (Core Spray or steam cooling with or without injection), these actions are designed to be an alternative method of providing adequate core cooling while actions are taken to reestablish core submergence. Sustained partial or total core uncovery can result in fuel damage and a significant release of fission products to the reactor coolant. Sustained core uncovery can also result in a breach of the RPV due to core melt material interaction with the RPV.

A Loss of core submergence will occur when the rate of inventory loss is greater than the rate of inventory makeup from high pressure injection sources. This condition can occur as the result of the following events/sequences:

A LOCA will cause RPV level to reach the Top of Active Fuel when the LOCA is the result of a large break (momentary core uncovery is expected to occur under this condition) or when the LOCA is due to a small or intermediate break in combination with an inability of high pressure injection sources to keep up with the leakrate.

A loss of high pressure injection sources without the presence of a LOCA will also result in RPV Level decreasing to TAF, due to continued reactor steam flow without makeup.

Either of these events/sequences results in a challenge to the Fuel Clad Barrier when RPV Level cannot be restored and maintained above the TAF due to core uncovery, hence classification at this threshold is appropriate. However, for both these sequences, low



EAL#: RB1.L

pressure ECCS are designed to inject to the RPV as RPV pressure decreases below the shutoff head of the pumps. RPV depressurization will occur either due to the LOCA or manual initiation of Emergency RPV Depressurization when RPV Level cannot be restored and maintained above -185 in., provided injection systems are available. This will allow for restoration of RPV Level and re-establishment of core submergence.

Until the wide range (WR) indicator is downscale or becomes questionable, it should be used to assure RPV water level is above TAF. Once WR is downscale, use compensated fuel zone indication (SPDS or graph) to assess RPV level above TAF and when compensated fuel zone level indication goes below TAF or becomes questionable, consider the threshold exceeded.

If RPV level cannot be determined, the indicated level value and trend have become unreliable to the extent that decisions concerning adequate core cooling cannot be made. RPV Flooding in accordance with EOP 206 or EOP 206A is required to assure continued core cooling. These EOPs specify actions to rapidly depressurize the RPV and establish sufficient RPV injection to cool the core by either steam cooling (with injection) or flooding the RPV to the main steam line penetrations. Entry to the RPV Flooding EOPs warrants classification of a SITE AREA EMERGENCY based on EAL FB1.P and this EAL.


1. NEI 99-01, Rev. 06, Table 9-F-2 RCS Barrier Loss 2.A 2. Technical Specifications Table 3.3.3-2 ECCS Actuation Instrumentation Setpoints 3. Technical Specifications Figure B3/4-3-1 Reactor Vessel water Level 4. HC.OP-EO.ZZ-0101(Q)-FC RPV Control 5. HC.OP-EO.ZZ-0206(Q)-FC RPV Flooding 6. HC.OP-EO.ZZ-0206A(Q)-FC ATWS - RPV Flooding 7. HC.OP-EO.ZZ-0101A(Q)-FC ATWS - RPV Control



EAL#: RB1.L




EAL#: RB2.L


Subcategory: PC Pressure




EAL:

Drywell pressure > 1.68 psig due to RCS leakage

Basis:

The 1.68 psig primary containment pressure is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the ECCS or equivalent makeup system.


The drywell high pressure scram setpoint (1.68 psig) is an entry condition to the EOP flowcharts 101 and 102. EOP Flowchart 102 will prescribe operation of drywell or suppression chamber sprays.

There are multiple Control Room indicators and alarms that can be used to determine the presence of this threshold. Overhead annunciators alarm at 1.0 psig, 1.5 psig and 1.68 psig. Plant automatic response to a high drywell pressure condition includes: a reactor scram, ECCS initiation, trip of the drywell cooling fans and isolation of the cooling water to the drywell. These actuations may mask the trend in drywell pressure. For example, the scram will result in less heat being added to the containment and the cooling water isolation will result in no heat being removed.

In the HCGS design basis, primary containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend. Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling.

The threshold phrase “…due to RCS leakage” focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect Drywell pressure. Drywell pressure greater than 1.68 psig with corollary indications (e.g., elevated drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.68 psig should not be considered an RCS barrier loss.



EAL#: RB2.L

Indication of an RCS leakage should be positively determined by observing drywell parameters, including drywell pressure and temperature trends, drywell equipment and floor drain sump levels, DAPA radiation levels, atmospheric pressure, torus pressure, and the status of drywell cooling systems.

An isolable Reactor Recirculation pump dual seal failure should not result in drywell pressure reaching this threshold, hence classification under this RCS Loss should not occur.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 5-F-2 RCS Barrier Loss 1.A 2. Technical Specifications Table 3.3.3-2 ECCS Actuation Instrumentation Setpoints 3. HC.OP-EO.ZZ-0101(Q)-FC Reactor/Pressure Vessel (RPV) Control 4. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 5. HC.OP-AB.CONT-0002(Q) Primary Containment 6. HC.OP-AB.CONT-0001(Q) Drywell Pressure 7. HC.OP-GP.ZZ-0005(Q) Drywell Leak Source Detection 8. HC.OP-SO.SM-0001(Q) Isolation Systems Operation



EAL#: RB3.L






EAL:

A break outside primary containment CANNOT be isolated from the Control Room in ANY of the following systems:

Main steam line

HPCI steam line

RCIC steam line

RWCU

Feedwater

Basis:

Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated from the Control Room, the RCS barrier Loss threshold is met.


A high energy line break that results in an isolation signal for any of the systems listed in the threshold requires closure of the associated Primary Containment Isolation valves to maintain RCS and Primary Containment integrity under abnormal conditions. A failure of these isolation valves to isolate allows reactor coolant to be released directly outside the Primary Containment (containment bypass), resulting in a Loss of RCS and Loss of Containment. A RCS line is any line that communicates directly with the reactor.



EAL#: RB3.L

This threshold allows for valve closure from the Control Room to isolate any high energy line not completely isolated, prior to event classification. Isolation is defined as the closure of any valve from the Control Room in the system(s) not completely isolated. For example, if the isolation logic fails to cause valve closure, but operator actions implemented in the Control Room successfully isolates the effected line(s), then event classification under this threshold is not warranted. This includes motor operated valves not controlled by the isolation logic, but that are controlled from the Control Room.

Although this threshold allows for valve closure from the Control Room, the time to attempt closure and make a decision if leak isolation was successful runs concurrently with the EAL 15-minute assessment clock:

If, during the EAL 15-minute assessment period attempts from the Control Room to isolate the leak are successful then, this threshold is NOT exceeded and classification per this threshold should NOT be made.

If, during the EAL 15-minute assessment period attempts from the Control Room to

isolate the leak are NOT successful then, this threshold is exceeded and classification should be made at that time.

If near the end of the 15 minute assessment period and the control room staff has not

been able to attempt leak isolation or the EC is not convinced that an isolation attempt has been successful, then this threshold is exceeded and classification should be made at or before the 15-minute assessment time expires.

Nuclear Steam Supply Shutoff System (NSSSS) isolations, as well as HPCI and RCIC steam line isolations, are associated with systems that are part of the RCS boundary and penetrate the Primary Containment. Isolation requirements for these lines are covered in 10CFR50, Appendix A, General Design Criteria 55. These systems form a closed loop outside the Primary Containment, and are not open or potentially open to the environment. They are included in this threshold since they represent an extension of the RCS boundary beyond the Primary Containment, and a potential release path from the RCS to the environment. Without complete isolation, continuing flow/leakage represents a situation where reactor coolant is discharging outside the Primary Containment, including areas in the Reactor Building addressed in the EOPs.

The isolation valve status of all isolation groups is monitored for quick reference on SPDS to be backed up by operator observation of valve status.

A high-energy line break may be detected by the following:

Main steam line: NSSSS ISLN SIG - STM TNL TEMP HI (C8-C4) NSSSS ISLN SIG - MN STM FLOW HI (C8-B4) MSIV CLOSURE (C5-B3)



EAL#: RB3.L

Rapid changes in Main Steam Line Flow and Steam Tunnel Temperatures

HPCI steam line: HPCI/RHR A AREA LEAK TEMP HI (D3-A1) HPCI STM LK ISLN TIMER INITIATED (D3-B1) HPCI STEAM LINE DIFF PRESSURE HI (B1-A5)

RCIC steam line: RCIC/RHR B AREA LEAK TEMP HI (D3-A2) RCIC STM LK ISLN TIMER INITIATED (D3-B2) RCIC STEAM LINE DIFF PRESSURE HI (B1-A2)

RWCU: MN STM/RWCU AREA LEAK TEMP HI (D3-A3) RWCU STM LK ISLN TIMER INITIATED (D3-B3) RWCU DIFF FLOW HI (C1-A2)

Feedwater line break in steam tunnel: NSSSS ISLN SIG - STM TNL TEMP HI (C8-C4) NSSSS ISLN SIG - MN STM FLOW HI (C8-B4) MSIV CLOSURE (C5-B3) Rapid changes in Feedwater Flow and Steam Tunnel Temperatures

Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an unisolable break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 RCS Barrier Loss 3.A 2. UFSAR 5.4.5 Main Steam Line Isolation System 3. UFSAR 6.3.1.1 Emergency Core Cooling System 4. UFSAR 5.4.6 RCIC 5. UFSAR 5.4.8 RWCU 6. UFSAR 5.4.9 Main Steam Line and Feedwater Line 7. HC.OP-AR.ZZ-0006, 8, 10, 12, 14 (Q) Annunciator Response Procedure, Windows

B1/C1/C5/C8/D3



EAL#: RB3.L

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EAL#: RB4.L






EAL:

Emergency RPV Depressurization is required

Basis:

Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs). Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.


The following EOPs require Emergency Depressurization under specific conditions: HC.OP-EO.ZZ-0101A-FC, ATWS - RPV Control

When RPV level cannot be restored and maintained above -185 in. HC.OP-EO.ZZ-0101-FC, Reactor/Pressure Vessel (RPV) Control

Before RPV level reaches -185 in. with available injection, or

When RPV level drops to -198 in. when in Steam Cooling



EAL#: RB4.L

HC.OP-EO.ZZ-0102-FC, Primary Containment Control

Cannot maintain Supp Chamber Pressure below Pressure Suppression Pressure Curve, or

Drywell Temperature cannot be restored and maintained below 340°F, or

Combined Supp Pool Temp and RPV Press cannot be maintained below the Heat Capacity Temperature Limit Curve, or

Supp Pool Level cannot be maintained above 38.5 in., or

Supp Pool Level and RPV Press cannot be restored and maintained below the SRV Tail Pipe level Limit Curve

HC.OP-EO.ZZ-0103/4-FC, Reactor Building & Rad Release Control

When selected Area Room Temperatures/Levels cannot be maintained below the Max Safe Operating Temperature/Floor Level and there is a Primary System discharging outside Primary Containment, or

When Gaseous Radioactive Release cannot be maintained below GE levels and there is a Primary System discharging outside Primary Containment and Reactor Building

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 RCS Barrier Loss 3.B 2. HC.OP-EO.ZZ-0101(Q)-FC Reactor/Pressure Vessel (RPV) Control 3. HC.OP-EO.ZZ-0101A(Q)-FC ATWS - RPV Control 4. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 5. HC.OP-EO.ZZ-0103/4(Q)-FC Reactor Building & Rad Release Control 6. EP FAQ 2015-001



EAL#: RB5.L






EAL:

ANY condition in the opinion of the Emergency Coordinator that indicates loss of the RCS barrier

Basis:

This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the RCS barrier is lost.


The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the RCS barrier is lost. Such a determination should include IMMINENT barrier degradation, barrier monitoring capability and dominant accident sequences.

Barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance.

Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators. This assessment should include instrumentation operability concerns, readings from portable instrumentation and consideration of offsite monitoring results.

Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.



EAL#: RB5.L

Definitions:


EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 RCS Loss 6.A



EAL#: CB1.P



Initiating Condition: Potential Loss of Containment


EAL# & Point Value: CB1.P (2 points)

EAL:

SAG entry is required

Basis:

The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold FB1.L. The Potential Loss requirement for entry into Severe Accident Guidelines indicates adequate core cooling cannot be assured and that core damage is possible. BWR EPGs/SAGs specify the conditions, when the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to assure adequate core cooling.

PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which, if not corrected, could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.


Core submergence is the mechanism of core cooling whereby each fuel element is completely covered with water. Indicated RPV water level at or above the top of the active fuel (TAF) constitutes the principal means of confirming the adequacy of core cooling via this mechanism. Assurance of continued adequate core cooling through core submergence is achieved when RPV water level can be maintained at or above TAF.



EAL#: CB1.P

Spray cooling is the mechanism of core cooling whereby the uncovered portion of the core is cooled by spray flow. Adequate spray cooling exists by design when at least one Core Spray loop is operating at design flow (6150 gpm) and RPV water level is at or above the elevation of the jet pump suctions (-215 inches). The covered portion of the core is then cooled by submergence while the uncovered portion is cooled by spray flow.

Steam cooling is the mechanism of core cooling whereby steam updraft up through the uncovered portion of each fuel bundle is sufficient to prevent the temperature of the hottest fuel rod from exceeding the appropriate limiting value, which is specific to the mode of steam cooling being employed (i.e., with and without injection of makeup water to the RPV).

With injection of makeup water into the RPV established, adequate core cooling exists when steam flow through the core is sufficient to preclude the peak clad temperature of the hottest fuel rod from exceeding 1500°F, the threshold temperature for fuel rod perforation. Assurance of continued adequate core cooling is achieved when RPV water level can be maintained at or above the Minimum Steam Cooling RPV Water Level -185 in.).

With no injection into the RPV established, adequate core cooling exists only so long as the covered portion of the reactor core generates sufficient steam to preclude the peak clad temperature of the hottest fuel rod from exceeding 1800°F, the threshold temperature for significant metal-water reaction.

Prolonged lack of cooling may result in severe overheating of the fuel clad, additional release of energy from accelerated clad oxidation, and eventual fuel melting. For events starting from full power operation, the failure to promptly reflood could result in some fuel melting. Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some time. Reactor Water Level remaining below TAF for an extended amount of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions.

Ample time should be allowed for Low Pressure ECCS and alternate injection systems to restore RPV level prior to entry into this classification. The time basis for deciding whether or not RPV level can be maintained > -185 in. should be based on the rate of reactor depressurization, the availability of low-pressure injection sources, (ECCS and alternate injection systems), and the rate of Reactor coolant inventory loss. Indications such as RPV level trend, injection flow rates, containment parameter trends, and low pressure injection system operability should also be considered.

In the event RPV level cannot be restored > -185 in., adequate core spray flow cannot be established or RPV water level cannot be determined with indication of fuel damage, SAG entry will be required by the EOPs.

This threshold is also a Loss of the Fuel Clad barrier (FB1.L). Since SAG entry occurs after core uncovery has occurred, a Loss of the RCS barrier exists (RB1.L). SAG entry, therefore, represents a Loss of two barriers and a Potential Loss of a third, which requires a GENERAL EMERGENCY classification.



EAL#: CB1.P

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Potential Loss 2.A 2. HC.OP-EO.ZZ-0101A(Q)-FC ATWS – RPV Control 3. HC.OP-EO.ZZ-0101(Q)-FC RPV Control 4. HC.OP-EO.ZZ-0206(Q)-FC RPV Flooding 5. HC.OP-EO.ZZ-0206A(Q)-FC ATWS - RPV Flooding 6. HC.OP-AM.ZZ-0001(Z) Severe Accident Guidelines (SAG) 7. EP-FAQ 2015-001



EAL#: CB1.P




EAL#: CB2.P


Subcategory: PC Conditions




EAL:

Drywell pressure > 62 psig

Basis:

The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.


If this threshold is exceeded, a challenge to the containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists. This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred.

EOP 102 requires intentional venting of the primary containment before drywell pressure reaches 65 psig (Primary Containment Pressure Limit). If the venting action is performed before this EAL threshold is exceeded, a Loss of the Containment Barrier exists under EAL CB4.L.


1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Potential Loss 1.A 2. UFSAR Table 6.2-1 Containment Design Parameters 3. UFSAR Section 6.2 Containment Systems 4. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control



EAL#: CB2.P




EAL#: CB3.P






EAL:

Indications of ≥ 6% H2 and ≥ 5% O2 in Drywell or Torus

Basis:

If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns, this threshold results in the declaration of a General Emergency.


Venting is required when Hydrogen concentration in the Drywell or Torus reaches the Explosive Mixture (deflagration concentrations) of ≥ 6% with Oxygen concentration ≥ 5%. Exceeding these parameters creates the potential for an UNISOLABLE breach of the primary containment, which could result in an uncontrolled, unmonitored, and untreated release of radioactivity to the environment. This EAL represents a Potential Loss of Containment, since containment venting is required due to Containment parameters potentially exceeding their design limits. Once the containment is vented per the EOPs, a loss of the Containment barrier has occurred and classification per EAL CB4.L should be made. The magnitude of any radiological release is dependent upon events leading to the requirement for emergency venting, including a loss of the RCS and a loss of the Fuel Clad Barriers.

The elevated hydrogen in the Drywell or Torus may result from excessive zircaloy-water reaction occurring following a LOCA. Additionally, hydrogen and oxygen gas may be introduced into the containment environment from long term disassociation of water in the Suppression Chamber.

SAG guidance in these cases is provided to vent the Primary Containment regardless of off-site dose consequences. Although radiological releases resulting from venting containment may exceed Environmental Protection Agency (EPA) Protective Action Guideline (PAG) limits,



EAL#: CB3.P

a controlled, monitored, and isolable release is preferred to a potential uncontrolled, unmonitored radiological release that would result from a failure of containment.

Elevated primary containment hydrogen concentration is alarmed at ≥ 2% by overhead annunciator E3-F5 (H2/O2 ANALYZER TROUBLE) and digital alarm points D4203 and D4212.

Definitions:


EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Potential Loss 1.B 2. HC.OP-AM.ZZ-0001(Z) (SAG-2) Containment and Radioactivity Release Control 3. HC.OP-AR.ZZ-0024(Q) CRIDS Computer Points Book 5 D3624 thru D4288 4. HC.OP-AR.ZZ-0016(Q) Overhead Annunciator Window Box E3



EAL#: CB4.P






EAL:

HCTL exceeded (EOP Curve SPT-P)

Basis:

The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

• Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized, OR

• Suppression chamber pressure above Primary Containment Pressure Limit A, while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.

The HCTL is a function of RPV pressure, suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore, the inability to maintain plant parameters below the limit constitutes a potential loss of containment.


The HCTL is given in EOP Flowchart 102 Curve SPT-P (ref. 2).

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Potential Loss 1.C 2. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 3. HC.OP-EO.ZZ-0106(Q)-FC Primary Containment Control HPCI/RCIC Only



EAL#: CB4.P



Hope Creek Page 1 of 2 Rev. 0)

EAL#: CB5.P


Subcategory: Radiation




EAL:

DAPA Radiation Monitor RI-4825A or RI-4825B reading EITHER of the following:

• With drywell sprays, ≥ 10,000 R/hr

• Without drywell sprays, ≥ 20,000 R/hr

Basis:

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss threshold. NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a GENERAL EMERGENCY.


Core damage analysis indicates 20% fuel damage dispersed into the drywell atmosphere is expected to produce a Drywell Atmosphere Post Accident (DAPA) radiation monitor reading of approximately 10,000 R/hr at 1 hr after shutdown (the most conservative) with drywell spray in service, and approximately 20,000 R/hr at 1 hr after shutdown with drywell spray not in service.

From Attachment 2 of HC.EP-EP.ZZ-0205, Step 6 the following formula was used to derive the EAL Set-points:

% Cladding Damage= �Indicated Radiation Level

100% Cladding Damage Radiation Level�×100


Hope Creek Page 2 of 2 Rev. 0)

EAL#: CB5.P

20 (% Cladding Damage) = DAPA Rad Monitor Reading X100 50,000R/HR (from Att 2. Fig. 1, of HC.EP-EP.ZZ-0205 with Sprays)

DAPA Reading = (20) (50,000R/HR ) 100

DAPA Reading = 10,000 R/hr (with Drywell Spray in service)

OR

20 (% Cladding Damage) = DAPA Rad Monitor Reading X100 100,000R/HR (from Att 2. Fig. 1, of HC.EP-EP.ZZ-0205 without Sprays)

DAPA Reading = (20) (100,000R/HR ) 100

DAPA Reading = 20,000 R/hr (without Drywell Spray in Service)

DAPA radiation monitors RE-4825A and RE-4825B are high range area radiation monitors and are located in the drywell at 145’ elevation. Each detector has a range of 1 to 108 R/hr and a built in check source that keeps the associated channel on-scale at approximately 1.2 R/hr. In the Control Room, DAPA radiation levels are displayed on RI-4825A (RM-11 9RX635) and B (RM-11 9RX636) ((Panel 10C604) and RR-4825A and B (10C650). Drywell radiation levels ≥ 2.0E+2 R/hr activates Overhead Annunciator C6-C1, RADIATION MONITORING ALARM/TRBL.

A structural I-beam in the drywell shields DAPA radiation monitor RE-4825B from shine due to radioactivity in the RCS piping. RE-4825A is not shielded. When the RCS is intact, elevated RCS coolant activity produces higher readings on RE-4825A than on RE-4825B. During events in which radioactivity is dispersed into the drywell atmosphere, structural shielding diminishes as the monitors become immersed in the radiation field. If the difference in RE-4825A and RE-4825B readings should decrease, it may be inferred that a LOCA event is occurring.

In order to facilitate the EAL declaration, the EAL threshold that specifies, “With Drywell Sprays ≥10,000R/Hr” should be used if Drywell Spray has been or is in service at any time during the event. It can be assumed that Drywell Spray in service for any amount of time results in some lowering of Drywell Radiation and therefore use of the lower threshold value is the correct EAL threshold to use.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Potential Loss 4.A 2. HC.EP-EP.ZZ-0205(Q) TSC – Post Accident Core Damage Assessment 3. UFSAR Table 11.5-1 Hope Creek Radiation Monitoring Systems 4. HC.OP-AR.ZZ-0011(Q) Overhead Annunciator Box C6



EAL#: CB6.P






EAL:

ANY condition in the opinion of the Emergency Coordinator that indicates potential loss of the Containment barrier

Basis:

This threshold addresses any other factors that may be used by the Emergency Coordinator in determining whether the Containment Barrier is potentially lost. The Emergency Coordinator should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.


The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Containment barrier is potentially lost. Such a determination should include IMMINENT barrier degradation, barrier monitoring capability and dominant accident sequences.




Definitions:



EAL#: CB6.P


EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Potential Loss 6.A



EAL#: CB1.L



Initiating Condition: Loss of Containment


EAL# & Point Value: CB1.L (3 points)

EAL:

UNPLANNED rapid drop in primary containment pressure following primary containment pressure rise

Basis:

Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of primary containment integrity. Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity. These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.


Conditions that result in a drop in primary containment (PC) pressure following a pressure rise that are not the direct result of a containment failure do not warrant classification under this threshold. These events include the initiation of drywell sprays, the re-establishment of drywell cooling, and anticipated drywell pressure drop due to ambient losses. Other indications such as Reactor Building Area Radiation Monitors (ARMs) radiation levels, Reactor Building area temperatures, Reactor Building floor and sump levels, Plant Effluent radiation levels, and containment isolation status should be used to confirm the loss of containment integrity if possible. Reactor Building to Torus vacuum breaker status should be monitored to ensure that this pathway does not result in a loss of containment integrity.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Barrier Loss 1.A 2. HC.OP-EO.ZZ-0101(Q)-FC Reactor/Pressure Vessel (RPV) Control



EAL#: CB1.L

3. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 4. HC.OP-AB.CONT-0002(Q) Primary Containment 5. HC.OP-AB.CONT-0001(Q) Drywell Pressure



EAL#: CB2.L






EAL:

Primary containment pressure response NOT consistent with LOCA conditions

Basis:

Primary containment pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity. These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.


This threshold relies on operator recognition of an unexpected response for the condition and, therefore, does not have a specific criterion. The unexpected response is important because it is the indicator for a containment bypass condition.

A downcomer failure, by itself, does not represent a loss of the Containment barrier. This failure, however, renders the primary containment inoperable per Technical Specifications because primary containment integrity has been compromised. A downcomer failure (bypass of the pressure suppression function) combined with a large break LOCA would likely result in a Potential Loss of the Containment barrier under Containment Potential Loss EAL CB2.P if Drywell pressure cannot be maintained below 62 psig.

UFSAR Section 6.2 provides a summary of primary containment pressure response for several postulated accident conditions resulting in the release of RCS inventory to the containment. These accidents include:

• Rupture of a recirculation line

• Rupture of a main steam line

• Intermediate size liquid line rupture



EAL#: CB2.L

• Small size RCS rupture

The primary containment pressure response to these breaks were bounded by the recirculation line break.

UFSAR Figures 6.2-3 and 6.2-7 (Attachment 2) illustrates the primary containment pressure response due to a recirculation line break. The maximum calculated drywell pressure is 50.6 psig and is well below the design allowable pressure of 62 psig.

Due to conservatisms in LOCA analyses, actual primary containment pressure response is expected to be less than the analyzed response. For example, blowdown mass flowrate may be only 60-80% of the analyzed rate, initial primary containment pressure may be less than 1.5 psig, etc.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Barrier Loss 1.B 2. UFSAR Section 6.2.1.1.3.1 Introduction (Pressure Suppression Containment Design

Evaluation) 3. UFSAR Figure 6.2-3 Short Term Containment Pressure Response Following Recirculation

Line Break 4. UFSAR Figure 6.2-7 Long Term Containment Pressure Response Following Recirculation

Line Break 5. UFSAR Table 6.2-1 Containment Design Parameters 6. UFSAR Table 6.2-3 Initial Conditions for Containment Response Analyses



EAL#: CB3.L


Subcategory: PC Isolation Failure




EAL:

UNISOLABLE direct downstream pathway to the environment exists after primary containment isolation signal

Basis:

This threshold addresses incomplete containment isolation that allows an UNISOLABLE direct release to the environment The use of the modifier “direct” in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.


This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the UNISOLABLE open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of primary containment integrity.



EAL#: CB3.L

As stated above, the adjective “Direct” modifies “release pathway” to discriminate against release paths through interfacing liquid systems. Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include UNISOLABLE main steam line or RCIC/HPCI steam line breaks, UNISOLABLE RWCU system breaks, and UNISOLABLE containment atmosphere vent paths. If the main condenser is available with an UNISOLABLE main steam line, there may be releases through the steam jet air ejectors and gland seal exhausters. These pathways are monitored, however, and do not meet the intent of an UNISOLABLE direct release path to the environment. These minor releases are assessed using EALs in Category R, Abnormal Rad Release / Rad Effluent.

Indications (symptoms) of primary containment failure may be evident without the exact pathway being understood at the time of the failure. If the primary containment or part of the RCS is required to be isolated and there are valid indications that the primary containment is not isolated, the Containment barrier should be considered lost.

Consistent with the definition of UNISOLABLE, this threshold allows for isolation actions both from the Control Room and locally to isolate any system not completely isolated, prior to event classification. This includes Motor Operated Valves not controlled by isolation logic, but that are controlled from the Control Room. For example, if the isolation logic fails to cause valve closure, but operator actions implemented in the Control Room successfully isolates the containment breach path, then classification under this threshold is NOT warranted.

Although this threshold allows for valve closure both locally and from the Control Room, the time to attempt closure and make a decision if containment leak isolation was successful runs concurrently with the EAL 15-minute assessment clock.

• If, during the EAL 15-minute assessment period attempts from the Control Room to isolate the containment are successful then, this threshold is NOT exceeded and classification per this threshold should NOT be made.

• If, during the EAL 15-minute assessment period attempts from both the Control Room

and locally to isolate the containment are NOT successful then, this EAL is exceeded and classification should be made at that time.

• If near the end of the 15-minute assessment period and the control room staff has not

been able to attempt containment isolation or the EC is not convinced that an isolation attempt has been successful, then this threshold is exceeded and classification should be made at or before the 15-minute assessment time expires.

Definitions:




EAL#: CB3.L

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Barrier Loss 3.A 2. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control



EAL#: CB3.L




EAL#: CB4.L






EAL:

Intentional primary containment venting per EOPs

Basis:

This threshold addresses incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.


Override DW/P-3, Step DW/P-18 or DW/P-21 of EOP 102, Primary Containment Control, may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded. The threshold is met when the operator begins venting the primary containment in accordance with EOP 318, not when actions are taken to bypass interlocks prior to opening the vent valves. Purge and vent actions specified in Step DW/P-1 of EOP 102 to control primary containment pressure below the drywell high pressure scram setpoint does not meet this threshold because such action is only permitted if offsite radioactivity release rates will remain below the ODCM limits.



EAL#: CB4.L

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Barrier Loss 3.B 2. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 3. HC.OP-EO.ZZ-318 (Q)-Containment Venting



EAL#: CB5.L






EAL:

UNISOLABLE primary system leakage that results in exceeding EITHER of the following:

• ANY EOP 103 Reactor Bldg room temperature Table 1, Column 2

• ANY Reactor Bldg rad level > 1000 times normal

EOP 103 TABLE 1

Area Description & Room NumberColumn 1

Max NormalOp Temp

CRD Pump Room (4202) 115°F 140°F

HPCI (4111)

Core Spray Pump Rooms A(4118) & C(4116)

RHR Pump Rooms A(4113) & C(4114)

SACS A & C (4309)

RCIC Pump Room (4110)

Core Spray Pump Rooms B(4104) & D(4105)

RHR Pump Rooms B(4109) & D(4107)

SACS B & D (4307)

RWCU Pipe Chase (4402)

115°F 250°F

115°F 140°F

115°F 140°F

115°F 140°F

115°F 250°F

115°F 140°F

115°F 140°F

115°F 140°F

160°F 160°F

Column 2Max Safe Op Temp

Basis:

This threshold addresses incomplete containment isolation that allows an UNISOLABLE direct release to the environment.

The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe



EAL#: CB5.L

shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.


The presence of elevated general area temperatures or radiation levels in the Reactor Building may be indicative of UNISOLABLE primary system leakage outside the primary containment. The maximum safe operating values define this Containment barrier threshold because they are indicative of problems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside primary containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in EOP 103 Reactor Building Control.

RB maximum safe operating temperatures are conservatively defined by the qualification temperature of safety related equipment in the area. The equipment qualification program has proven that safety related equipment will perform satisfactorily to at least this temperature. In an area with multiple components and different qualification temperatures, the maximum safe operating temperature assigned to that area is generally the lowest of the individual temperatures.

The Maximum Safe Operating Radiation Level has been determined to be any area dose rate that increases by a factor of 1000. This value can be readily determined using the trend displays of the Radiation Monitoring System RM-11 computer. These displays can be monitored from the Control Room, Radiological Control Point, Upper Relay Room, and Technical Support Center.

In general, multiple indications should be used to determine if a primary system is discharging outside primary containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.

This threshold allows for valve closure to isolate any systems not completely isolated, prior to event classification. Isolation is defined as the closure of any valve in the system(s) not completely isolated. For example, if the isolation logic fails to cause valve closure, but operator actions successfully isolates the containment breach/leak path, then classification under this threshold is not warranted. This includes manual and motor operated valves either locally or remotely.



EAL#: CB5.L

Although this EAL allows for valve closure both remotely and locally, the time to attempt closure and make a decision if leak isolation was successful runs concurrently with the EAL 15-minute assessment clock:

• If, during the EAL 15-minute assessment period attempts to isolate the leak are successful then, this threshold is NOT exceeded and classification per this threshold should NOT be made.

• If, during the EAL 15-minute assessment period attempts to isolate the leak are NOT successful then, this threshold is exceeded and classification should be made at that time.

• If near the end of the 15 minute assessment period and the operators have not been able to attempt leak isolation or the EC is not convinced that an isolation attempt has been or will be successful, then this threshold is exceeded and classification should be made at or before the 15-minute assessment time expires.

Definitions:


EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Barrier Loss 3.C 2. HC.OP-EO.ZZ-0103/4(Q)-FC Reactor Building and Rad Release Control



EAL#: CB5.L




EAL#: CB6.L






EAL:

ANY condition in the opinion of the Emergency Coordinator that indicates loss of the Containment barrier

Basis:

This threshold addresses any other factors that are to be used by the Emergency Coordinator in determining whether the Containment barrier is lost.


The Emergency Coordinator judgment threshold addresses any other factors relevant to determining if the Containment barrier is lost. Such a determination should include IMMINENT barrier degradation, barrier monitoring capability and dominant accident sequences.




Definitions:




EAL#: CB6.L

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, Table 9-F-2 Containment Loss 6.A



EAL#: CU1.1

EAL Category: C – Cold Shutdown / Refuel System Malfunction

EAL Subcategory: 1 – RPV Level

Initiating Condition: UNPLANNED loss of RPV inventory

OPCON Applicability: 4 - Cold Shutdown, 5 - Refueling

EAL# & Classification Level: CU1.1 – UNUSUAL EVENT

EAL:

UNPLANNED loss of reactor coolant results in RPV water level less than a required lower limit for ≥ 15 min. (Note 1)


Basis:

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered. This EAL recognizes that the minimum required RPV level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document. The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.


RPV water level less than a required lower limit is meant to be less than the lower end of the level control band being procedurally maintained for the current condition or evolution.



EAL#: CU1.1

RPV water level is normally monitored using the instrument ranges in Attachment 2.

In preparation for RPV floodup and refueling operations, the shutdown range instrument transmitter LT-N027 is valved out and replaced with LT-11683. The shutdown range indication is thus expanded to 0” - 550”.

Definitions:


EAL Basis Reference(s): 1. NEI 99-01, Rev. 06, CU1 Example EAL #1 2. Technical Specifications Table 2.2.1-1 Reactor Protection System Instrumentation

Setpoints 3. Technical Specifications Figure B3/4-3-1, Reactor Vessel Water Level 4. HC.OP-IO.ZZ-0005(Q) Cold Shutdown to Refueling 5. HC.IC-GP.BB-0003(Q) Nuclear Boiler - Nondivisional Channel LT-11683 / B21-N027 Rx

Cavity Flood Up Level / Rx Shutdown Range Level Setup



EAL#: CU1.2



Initiating Condition: UNPLANNED loss of RCS inventory



EAL:

RPV water level CANNOT be monitored AND

UNPLANNED increase in ANY Table C-1 sump or tank levels due to a loss of RPV inventory

Table C-1 Sumps & Tanks

Drywell equipment drain sump

Drywell floor drain sump

Reactor Building equipment drain sump

Reactor Building floor drain sump

Suppression Pool Observation of RCS leakage that is UNISOLABLE

Basis:

This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an UNUSUAL EVENT due to the reduced water inventory that is available to keep the core covered. This EAL addresses a condition where all means to determine RPV level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated



EAL#: CU1.2

against other potential sources of water flow to ensure they are indicative of leakage from the RCS. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.


In preparation for RPV floodup and refueling operations, the shutdown range instrument transmitter LT-N027 is valved out and replaced with LT-11683. The shutdown range indication is thus expanded to 0” - 550”. During refueling operations, visual observation by personnel on the refuel floor in communication with the Control Room may also provide indication of reactor cavity water level and RPV water level.

In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by increases in sumps and tanks listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV. A Reactor Building equipment or floor drain sump level rise may also be indicative of RPV inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool level could be indicative of RHR valve misalignment or leakage. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory.

Definitions:


EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, CU1 Example EAL #2 2. HC.IC-GP.BB-0003(Q) Nuclear Boiler - Nondivisional Channel LT-11683 / B21-N027 Rx

Cavity Flood Up Level / Rx Shutdown Range Level Setup 3. UFSAR 5.2.5 Detection of Leakage Through the Reactor Coolant Pressure Boundary and

from the Emergency Core Cooling System 4. UFSAR 7.6.1.3 Leak Detection System – Instrumentation and Controls 5. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 6. HC.OP-GP.ZZ-0005(Q) Drywell Leak Source Detection



EAL#: CA1.1



Initiating Condition: Significant loss of RPV inventory


EAL# & Classification Level: CA1.1 – ALERT

EAL:

Loss of RPV inventory as indicated by compensated RPV water level < -38 in.

Basis:

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For this EAL, a lowering of RPV water level below -38 in. indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery. Although related, this EAL is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Residual Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.

If RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.


The threshold RPV level of -38 in. is the low-low high pressure ECCS actuation setpoint. RPV level is normally monitored using compensated level as read directly from SPDS or the compensation curves in the IOs.



EAL#: CA1.1

EAL Basis Reference(s):

1. NEI 99-01, Rev. 06, CA1 Example EAL #1 2. HCGS Technical Specifications Table 3.3.3-2, ECCS Actuation Instrumentation Setpoints 3. HCGS Technical Specifications Figure B3/4-3-1 Reactor Vessel Water Level 4. HC.OP-IO.ZZ-0005(Q) Cold Shutdown to Refueling 5. HC.OP-IO.ZZ-0009 (Q) Refueling Operations 6. HC.OP-AB.RPV-0009(Q) Shutdown Cooling



EAL#: CA1.2



Initiating Condition: Loss of RPV inventory



EAL:

RPV water level CANNOT be monitored for ≥ 15 min. (Note 1) AND

UNPLANNED increase in ANY Table C-1 sump or tank levels due to a loss of RPV inventory








Basis:

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety. For this EAL, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV.



EAL#: CA1.2

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.

If the RPV inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.


In preparation for RPV floodup and refueling operations, the shutdown range instrument transmitter LT-N027 is valved out and replaced with LT-11683. The shutdown range indication is thus expanded to 0” - 550”. During refueling operations, visual observation by personnel on the refuel floor in communication with the Control Room may also provide indication of reactor cavity water level and RPV water level.

In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by increases in sumps and tanks listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV. A Reactor Building equipment or floor drain sump level rise may also be indicative of RPV inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool level could be indicative of RHR valve misalignment or leakage. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory.

Definitions:



EAL Basis Reference(s): 1. NEI 99-01, Rev. 06, CA1 Example EAL #2 2. HC.IC-GP.BB-0003(Q) Nuclear Boiler - Nondivisional Channel LT-11683 / B21-N027 Rx

Cavity Flood Up Level / Rx Shutdown Range Level Setup 3. UFSAR 5.2.5 Detection of Leakage Through the Reactor Coolant Pressure Boundary and

from the Emergency Core Cooling System 4. UFSAR 7.6.1.3 Leak Detection System – Instrumentation and Controls 5. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 6. HC.OP-GP.ZZ-0005(Q) Drywell Leak Source Detection



EAL#: CS1.1



Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability


EAL# & Classification Level: CS1.1 – SITE AREA EMERGENCY

EAL:

CONTAINMENT CLOSURE NOT established AND RPV compensated level < -129 in.

OR

CONTAINMENT CLOSURE established AND RPV compensated level < -161 in.

Basis:

This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a SITE AREA EMERGENCY declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV water level cannot be restored, fuel damage is probable. Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RPV inventory control functions. The difference in the specified RPV levels of the first and second condition reflect the fact that with CONTAINMENT CLOSURE established, there is a lower probability of a fission product release to the environment. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1




EAL#: CS1.1

When RPV level decreases to -129 in., RPV water level is at the low pressure (low-low-low) ECCS actuation setpoint. When RPV level drops to the top of active fuel (an indicated RPV level of -161 in.), core uncovery starts to occur. RPV level is normally monitored using compensated level as read directly from SPDS or the compensation curves in the IOs.

The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier.

Sustained core uncovery can result in core damage and a significant release of fission products to the RCS and could ultimately challenge the Containment barrier.

CONTAINMENT CLOSURE can be established by meeting the requirements of any of the following:

OP-HC-108-115-1001, Operability Assessment and Equipment Control Program

Technical Specifications 3.6.5.1, Secondary Containment Integrity

Technical Specifications 3.6.1.1, Primary Containment Integrity

Definitions:


CONTAINMENT CLOSURE: Is the procedurally defined actions taken to secure the Containment (Primary or Secondary) and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

EAL Basis Reference(s): 1. NEI 99-01, Rev. 06, CS1 Example EAL #1 & #2 2. OU-AA-103 Shutdown Safety Management Program 3. HCGS Technical Specifications Table 3.3.3-2, ECCS Actuation Instrumentation Setpoints 4. HCGS Technical Specifications Figure B3/4-3-1, Reactor Vessel Water Level 5. HC.OP-AM-ZZ-0001(Z) Severe Accident Guidelines 6. OP-HC-108-115-1001 Operability Assessment and Equipment Control Program 7. Technical Specifications 3.6.5.1 Secondary Containment Integrity 8. Technical Specifications 3.6.1.1 Primary Containment Integrity



EAL#: CS1.2



Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability


EAL# & Classification Level: CS1.2 – SITE AREA EMERGENCY

EAL:


Core uncovery is indicated by UNPLANNED increase in ANY Table C-1 sump or tank levels of sufficient magnitude to indicate core uncovery








Basis:

This IC addresses a significant and prolonged loss of RPV inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a SITE AREA EMERGENCY declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and



EAL#: CS1.2

plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CG1 or RG1


This EAL applies to conditions in which the loss of decay heat removal capability has caused a significant drop in RPV water level and core uncovery may be challenged. If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected by increasing sump or tank level indications listed in Table C-1.

Sumps and tanks that may indicate an increased level due to RPV leakage are listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV. A Reactor Building equipment or floor drain sump level rise may also be indicative of RPV inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool level could be indicative of RHR valve misalignment or leakage. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory.

Definitions:






EAL#: CS1.2

EAL Basis Reference(s): 1. NEI 99-01, Rev. 06, CS1 Example EAL #3 2. Technical Specifications 3.9.2, Instrumentation – Refueling Operations 3. HC.OP-AR.ZZ-0019, Overhead Annunciator Window Box E6 4. UFSAR 5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary and

From the Emergency Core Cooling System 5. UFSAR 7.6.1.3 Leak Detection System - Instrumentation and Controls 6. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 8. HC.OP-GP.ZZ-0005(Q) Drywell Leak Source Detection



EAL#: CS1.2




EAL#: CG1.1



Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged


EAL# & Classification Level: CG1.1 – GENERAL EMERGENCY

EAL:

RPV compensated level < -161 in. for ≥ 30 min. (Note 1)

AND

ANY Containment Challenge indication, Table C-2


Note 9: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-min. time limit, declaration of a GENERAL EMERGENCY is NOT required.

Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE NOT established (Note 9)


UNPLANNED rise in drywell pressure

ANY Reactor Bldg rad level > 1000 times normal

Basis:

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.



EAL#: CG1.1

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.


When RPV compensated level drops to the top of active fuel (an indicated RPV level of -161 in.), core uncovery starts to occur. Four conditions are associated with a challenge to containment:

The status of CONTAINMENT CLOSURE indicates the ability to rely on the Primary or Secondary Containment as a barrier to fission product release. CONTAINMENT CLOSURE can be established by meeting the requirements of any of the following:




Explosive (deflagration) mixtures in the Drywell or Torus are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of ≥ 6%



EAL#: CG1.1

and oxygen concentration of ≥ 5% are considered the global deflagration concentration limits (ref. 4). Elevated Containment hydrogen concentration is alarmed at ≥ 2% by overhead annunciator E3-F5, H2/O2 ANALYZER TROUBLE, and digital alarm points D4203 and D4212.

An UNPLANNED rise in Drywell pressure in the Cold Shutdown or Refueling OPCON may signify an energy addition to the Primary Containment such that the Primary Containment cannot be relied upon as a barrier to fission product release.

Reactor Building area radiation monitors should provide indication of increased release that may be indicative of a challenge to CONTAINMENT CLOSURE. The Maximum Safe Operating radiation level is 1000 times normal and is indicative of problems in the Secondary Containment that may be spreading.

The Table C-2 Containment Challenge Indications applicable in the Refueling OPCON may likely consist only of the status of Secondary Containment and Reactor Building radiation levels. The Primary Containment is opened to the Secondary Containment in this OPCON and it is unlikely, if not impossible, to detect hydrogen concentrations in the deflagration range or rising Drywell pressure indications. Definitions:

CONTAINMENT CLOSURE: Is the procedurally defined action taken to secure the Containment (Primary or Secondary) and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.



EAL Basis Reference(s): 1. NEI 99-01, Rev. 06, CG1 Example EAL #1 2. OU-AA-103 Shutdown Safety Management Program 3. HCGS Technical Specifications Table 3.3.3-2 ECCS Actuation Instrumentation Setpoints 4. HC.OP-AM-ZZ-0001(Z) Severe Accident Guidelines 5. HC.OP-AM.ZZ-0001 (SAG-2) Containment and Radioactivity Release Control 6. HC.OP-AR.ZZ-0024(Q) CRIDS Computer Points Book 5 D3624 thru D4288 7. HC.OP-AR.ZZ-0016(Q) Overhead Annunciator Window Box E3 8. HC.OP-EO.ZZ-0103/4(Q)-FC Reactor Building and Rad Release Control 9. Technical Specifications 3.6.5.1 Secondary Containment Integrity 10. Technical Specifications 3.6.1.1 Primary Containment Integrity



EAL#: CG1.1




EAL#: CG1.2



Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with Containment challenged


EAL# & Classification Level: CG1.2 – GENERAL EMERGENCY

EAL:


Core uncovery is indicated by UNPLANNED increase in ANY Table C-1 sump or tank levels of sufficient magnitude to indicate core uncovery

AND ANY Containment Challenge indication, Table C-2


Note 9: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-min. time limit, declaration of a GENERAL EMERGENCY is NOT required.









EAL#: CG1.2

Table C-2 Containment Challenge Indications

CONTAINMENT CLOSURE NOT established (Note 9)


UNPLANNED rise in drywell pressure

ANY Reactor Bldg rad level > 1000 times normal

Basis: This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable. With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required. The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.



EAL#: CG1.2

The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV. This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.


This EAL applies to conditions in which the loss of decay heat removal capability has caused a significant drop in RPV water level and core uncovery may be challenged. If RPV level monitoring capability is unavailable, the RPV inventory loss must be detected increasing sump or tank level indications listed in Table C-1.

Sumps and tanks that may indicate an increased level due to RPV leakage are listed in Table C-1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Drywell equipment and floor drain sump level rise is the normal method of monitoring and calculating leakage from the RPV. A Reactor Building equipment or floor drain sump level rise may also be indicative of RPV inventory losses external to the Primary Containment from systems connected to the RPV. With RHR System operating in the Shutdown Cooling mode, an unexplained rise in suppression pool level could be indicative of RHR valve misalignment or leakage. Visual observation of leakage from systems connected to the RCS in areas outside the Primary Containment that cannot be isolated could be indicative of a loss of RPV inventory.

Four conditions are associated with a challenge to Containment:

The status of CONTAINMENT CLOSURE indicates the ability to rely on the Primary or Secondary Containment as a barrier to fission product release. CONTAINMENT CLOSURE can be established by meeting the requirements of any of the following:




Explosive (deflagration) mixtures in the Drywell or Torus are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by



EAL#: CG1.2

metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of ≥ 6% and oxygen concentration of ≥ 5% are considered the global deflagration concentration limits. Elevated Containment hydrogen concentration is alarmed at ≥ 2% by overhead annunciator E3-F5 (H2/O2 ANALYZER TROUBLE) and digital alarm points D4203 and D4212.

An UNPLANNED rise in Drywell pressure in the Cold Shutdown or Refueling OPCON may signify an energy addition to the Primary Containment such that the Primary Containment cannot be relied upon as a barrier to fission product release.

Reactor Building area radiation monitors should provide indication of increased release that may be indicative of a challenge to CONTAINMENT CLOSURE. The Maximum Safe Operating radiation level is 1000 times normal and is indicative of problems in the Secondary Containment that may be spreading.

The Table C-2 Containment Challenge Indications applicable in the Refueling OPCON may likely consist only of the status of Secondary Containment and Reactor Building radiation levels. The Primary Containment is opened to the Secondary Containment in this OPCON and it is unlikely, if not impossible, to detect hydrogen concentrations in the deflagration range or rising Drywell pressure indications.

Definitions:

UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally. CONTAINMENT CLOSURE: Is the procedurally defined actions taken to secure the Containment (Primary or Secondary) and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown. IMMINENT: The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.



EAL#: CG1.2

EAL Basis Reference(s): 1. NEI 99-01, Rev. 06, CG1 Example EAL #2 2. OU-AA-103 Shutdown Safety Management Program 3. Technical Specifications 3.9.2 Instrumentation – Refueling Operations 4. HC.OP-AR.ZZ-0019, Overhead Annunciator Window Box E6 5. UFSAR 5.2.5 Detection of Leakage Through Reactor Coolant Pressure Boundary and

From the Emergency Core Cooling System 6. UFSAR 7.6.1.3 Leak Detection System - Instrumentation and Controls 7. HC.OP-EO.ZZ-0102(Q)-FC Primary Containment Control 8. HC.OP-GP.ZZ-0005(Q) Drywell Leak Source Detection 9. HC.OP-AM.ZZ-0001 (SAG-2) Containment and Radioactivity Release Control 10. HC.OP-AR.ZZ-0024(Q) CRIDS Computer Points Book 5 D3624 thru D4288 11. HC.OP-AR.ZZ-0016(Q) Overhead Annunciator Window Box E3 12. HC.OP-EO.ZZ-0103/4(Q)-FC Reactor Building and Rad Release Control 13. Technical Specifications 3.6.5.1 Secondary Containment Integrity 14. Technical Specifications 3.6.1.1 Primary Containment Integrity



EAL#: CG1.2




EAL#: CU2.1



Initiating Condition: Loss of ALL but one AC power source to vital buses for 15 minutes or longer

OPCON Applicability: 4 - Cold Shutdown, 5 – Refueling, D - Defueled


EAL:

AC power capability to 4.16 KV vital buses reduced to a single power source for ≥ 15 min. (Note 1)

AND ANY additional single power source failure will result in loss of ALL AC power to SAFETY SYSTEMS


Basis:

This EAL describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an ALERT because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant. An “AC power source” is a source recognized in AOPs and EOPs, and capable of supplying required power to a vital bus. Some examples of this condition are presented below.

A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of vital buses being back-fed from the unit main generator.

A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being fed from an offsite power source.



EAL#: CU2.1

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.


“Capability” means the power source can be aligned to provide power to a vital bus within 15 minutes or is currently supplying power to at least one vital bus.


Emergency Classification escalates to an ALERT under EAL CA1.1 based on a loss of all offsite and all onsite AC power to all 4.16 KV vital buses.

This cold condition UNUSUAL EVENT EAL is equivalent to the hot condition ALERT EAL SA1.1.

Definitions:




EAL Bases Reference(s): 1. NEI 99-01, Rev. 06 – CU2 Example EAL #1 2. 205415-A-8765 Sheet 1 500 kv Transmission Plan & Profiles 3. E-0001-0 sheet 1 Hope Creek Generating Station Single Line Diagram 4. UFSAR 8.1.2 Onsite Power Systems 5. HCGS Technical Specifications 3.8.1.2, A.C. Sources-Shutdown



EAL#: CA2.1



Initiating Condition: Loss of ALL offsite and ALL onsite AC power to vital buses for 15 minutes or longer

OPCON Applicability: 4 - Cold Shutdown, 5 - Refueling, D - Defueled


EAL:

Loss of ALL offsite and ALL onsite AC power to 4.16 KV vital buses for ≥ 15 min. (Note 1)


Basis:

This EAL addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore a vital bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via IC CS1 or RS1.


The intent of this EAL is to classify degraded AC power events that result in a loss of all offsite power sources 13.8 KV to the 4.16 KV vital buses along with a loss of all onsite power sources (EDGs).


This cold condition ALERT EAL is equivalent to the hot condition SITE AREA EMERGENCY EAL SS1.1.



EAL#: CA2.1

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06 – CA2 Example EAL #1 2. HC.OP-AB.ZZ-0135 (Q) Station Blackout / Loss of Offsite Power / Diesel Generator

Malfunction 3. 205415-A-8765 Sheet 1 500 kv Transmission Plan & Profiles 4. E-0001-0 sheet 1 Hope Creek Generating Station Single Line Diagram 5. UFSAR 8.1.2 Onsite Power Systems 6. HCGS Technical Specifications 3.8.1.2, A.C. Sources-Shutdown



EAL#: CU3.1


EAL Subcategory: 3 – RCS Temperature

Initiating Condition: UNPLANNED increase in RCS temperature



EAL:

UNPLANNED increase in RCS temperature to > 200°F

Basis:

This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3. In the absence of reliable RCS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU3.2 should RCS level indication be subsequently lost.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.


The Technical Specification cold shutdown temperature limit is 200°F.



EAL#: CU3.1

RCS coolant temperature may be indicated by the following instrumentation:

Recirc pump suction loop A and B temperature on TR-R650-B31 (10C650C)

Recirc loop temperature from the following computer points: o Recirc loop A inlet temp 1 A221 o Recirc loop A inlet temp 2 A222 o Recirc loop B inlet temp 1 A223 o Recirc loop B Inlet temp 2 A224 o Recirc loop A avg inlet temp B2042 o Recirc loop B avg inlet temp B2043

RHR A & B Hx inlet temperatures from computer points A2380 & A2382

RWCU bottom head drain temperature from computer point A2942

Definitions:


EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, CU3 Example EAL #1 2. Technical Specifications Table 1.2, Operational Conditions 3. HC.OP-IO.ZZ-0004(Q) Shutdown from Rated Power to Cold Shutdown 4. HC.OP-SO.BC-0002 (Q) Decay Heat Removal Operation 5. HC.OP-IO.ZZ-0005(Q) Cold Shutdown to Refueling



EAL#: CU3.2



Initiating Condition: UNPLANNED increase in RCS temperature



EAL:

Loss of ALL RCS temperature and RPV water level indication for ≥ 15 min. (Note 1)


Basis:

This EAL addresses the inability to determine RCS temperature and level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Coordinator should also refer to IC CA3. This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.


RPV water level is normally monitored using the instrument ranges in Attachment 2. In preparation for RPV floodup and refueling operations, the shutdown range instrument transmitter LT-N027 is valved out and replaced with LT-11683. The shutdown range indication is thus expanded to 0” - 550”. During refueling operations, visual observation by personnel on the refuel floor in communication with the Control Room may also provide indication of reactor cavity water level and RPV water level. RCS coolant temperature may be indicated by the following instrumentation:



EAL#: CU3.2




RWCU bottom head drain temperature from computer point A2942

Definitions:




EAL#: CU3.2

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, CU3 Example EAL #2 2. Technical Specifications Figure B3/4-3-1, Reactor Vessel Water Level 3. HC.OP-IO.ZZ-0005(Q) Cold Shutdown to Refueling 4. HC.IC-GP.BB-0003(Q) Nuclear Boiler - Nondivisional Channel LT-11683 / B21-N027 Rx

Cavity Flood Up Level / Rx Shutdown Range Level Setup 5. HC.OP-IO.ZZ-0004(Q) Shutdown from Rated Power to Cold Shutdown 6. HC.OP-SO.BC-0002 (Q) Decay Heat Removal Operation



EAL#: CU3.2




EAL#: CA3.1



Initiating Condition: Inability to maintain plant in cold shutdown



EAL:

UNPLANNED increase in RCS temperature to > 200°F for > Table C-3 duration (Note 1)

OR

UNPLANNED RCS pressure increase > 10 psig

Table C-3 RCS Heat-up Duration Thresholds

RCS Status CONTAINMENT CLOSURE Status Heat-up Duration

Intact NOT Applicable 60 minutes **

NOT Intact Established 20 minutes **

NOT Intact NOT Established 0 minutes

** If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is NOT applicable

Basis:

This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant. In the absence of reliable RCS temperature indication caused by the loss of decay heat removal capability, classification should be based on the RCS pressure increase criteria when in Mode 4 (RCS intact) or based on time to boil data when in Mode 5 or RCS is not intact.



EAL#: CA3.1

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact.). The 20-minute criterion was included to allow time for operator action to address the temperature increase. The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Finally, in the case where there is an increase in RCS temperature, the RCS is not intact , and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top of irradiated fuel. The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability. Escalation of the emergency classification level would be via IC CS1 or RS1.


200°F is the Technical Specification cold shutdown temperature limit. A 10 psig RCS pressure increase is readable in the Control Room on PI-5824A (0 - 50 psig) and PI-5824B (0 - 50 psig). RCS coolant temperature may be indicated by the following instrumentation:




RWCU bottom head drain temperature from computer point A2942 CONTAINMENT CLOSURE can be established by meeting the requirements of any of the following:



EAL#: CA3.1

OP-HC-108-115-1001, Operability Assessment and Equipment Control Program, Attachment 5



Definitions:


CONTAINMENT CLOSURE: Is the procedurally defined actions taken to secure the Containment (Primary or Secondary) and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, CA3 Example EAL #1 & #2 2. OU-AA-103 Shutdown Safety Management Program 3. HCGS Technical Specifications Table 1.2, Operational Conditions 4. HC.OP-AB.RPV-0009(Q) Shutdown Cooling 5. HC.OP-IO.ZZ-0004(Q) Shutdown from Rated Power to Cold Shutdown 6. HC.OP-SO.BC-0002 (Q), Decay Heat Removal Operation 7. Technical Specifications 3.6.5.1 Secondary Containment Integrity 8. Technical Specifications 3.6.1.1 Primary Containment Integrity



EAL#: CA3.1




EAL#: CU4.1



Initiating Condition: Loss of Vital DC power for 15 minutes or longer

OPCON Applicability: 4 - Cold Shutdown, 5 – Refueling


EAL:

Loss of ANY of the following required vital 125 V DC Power Channel combinations as indicated by voltage < 108 V DC for ≥ 15 min. (Note 1):

Channel A and Channel B

Channel A, Channel C (either bus) and Channel D (either bus)

Channel B, Channel C (either bus) and Channel D (either bus)


Basis:

This IC addresses a loss of vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant. As used in this EAL, “required” means the vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event, escalation of the emergency classification level would be via IC CA1 or CA3, or an IC in Recognition Category R.



EAL#: CU4.1



125 VDC Power Channels A-D provide control power to Engineered Safety Features actuation, diesel generator auxiliaries, plant alarm and indication circuits, as well as the control power for the associated loads. If 125 VDC power is lost for an extended period of time (greater than 15 minutes) critical plant functions such as 4.16 KV breaker controls, CS and RHR pump controls required to maintain safe plant conditions may not operate, and core uncovery with subsequent Reactor Coolant System (RCS) and Primary Containment failure might occur. Both the RCS and Primary Containment may already be open if in the Refueling mode.

In Operational Condition 4 or 5, a minimum of two of the four 125 VDC Power Channels are required by Technical Specifications, including either Channel A (10D410) or Channel B (10D420). With the loss of three of the Channels (two if both Channels A and B are lost) core alterations are to be suspended and handling of recently irradiated fuel in the Secondary Containment and operations with a potential for draining the reactor vessel are to be stopped.

This UNUSUAL EVENT EAL is the cold condition equivalent of the hot condition loss of DC power SITE AREA EMERGENCY EAL SS2.1.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, CU4 Example EAL #1 2. Calculation E-4.1(Q) HC Class 1E 125 VDC Station Battery and Charger Sizing 3. UFSAR 8.3.2 DC Power Systems 4. HCGS Technical Specifications 3.8.2.2, DC Sources - Shutdown 5. HC.OP-AB.ZZ-0135 (Q) Station Blackout /Loss of Offsite Power/Diesel Generator

Malfunction, Table 5.3 Battery Bank Capacities



EAL#: CU5.1

EAL Category: C – Cold Shutdown / Refueling System Malfunction

EAL Subcategory: 5 – Loss of Communications

Initiating Condition: Loss of ALL onsite or offsite communications capabilities

OPCON Applicability: 4 - Cold Shutdown, 5 - Refueling, D – Defueled


EAL:

Loss of ALL Table C-4 onsite communication methods OR

Loss of ALL Table C-4 offsite communication methods OR

Loss of ALL Table C-4 NRC communication methods

Table C-4 Communication Methods

System Onsite Offsite NRC

Direct Inward Dial System (DID) X X X

Station Page System (Gaitronics) X

Station Radio System X

Nuclear Emergency Telephone System (NETS) X X

Centrex Phone System (ESSX) X X

NRC (ENS) X



EAL#: CU5.1

Basis:

This EAL addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to Offsite Response Organizations (OROs) and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant, privately owned equipment, relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations, etc.). The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations. The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are the States of New Jersey and Delaware. The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.


Onsite, offsite and NRC communications include one or more of the systems listed in Table C-4.

Direct Inward Dial System (DID)

Direct Inward Dial (DID) system is named for the dominant feature of the commercial telephone service provided by the local telephone company for the site. DID allows station telephones to be extensions or tied lines of the same systems. These exchanges can take advantage of backup power supplies provided to the stations, and may use either PSEG microwave, commercial telephone system microwave, or buried cable transmission systems to maintain external communications. This commercial telephone service is available as an additional backup for the NETS and Centrex/ESSX 1 system.

Station Page System (Gaitronics)

Gaitronics is a completely transistorized voice communication system with five voice channels: one page and five party. The system is designed for use in extreme environmental conditions such as dust, moisture, heat and noise. The system consists of handsets, speakers and their associated amplifiers. The power for this system is 120 volts AC from an inverted DC source to provide reliable communications during an emergency.

Station Radio System



EAL#: CU5.1

The Operations and Fire Protection Department UHF radio system is a multi-frequency system used routinely by both station Operations Departments and the Fire Protection Department. When an emergency event is declared, these radio frequencies serve both station Operations Support Centers (OSC).

Nuclear Emergency Telephone System (NETS)

The Nuclear Emergency Telecommunications System (NETS) is a privately controlled, self-contained telephone exchange that operates as a closed system, not accessible from other phone exchanges. This feature allows the system to be dedicated to emergency response use. The system may use PSEG microwave, commercial telephone system microwave, fiber optics, or buried cable transmission as needed. The exchange switching equipment is maintained at the Environmental & Energy Resource Center (EERC). As an independent system with an uninterruptible power supply, it may operate with or without local phone service or external power.

Centrex Phone System (ESSX)

The Centrex/Electronic Switch System Exchange 1 (Centrex/ESSX 1) is also a privately controlled exchange, which PSEG operates with its own microwave signal system. This system is also independent of local phone service, since each circuit is independently wired. The microwave signal is generated from corporate facilities in Newark, NJ, separated from any local effects of weather or telephone use. The exchange is accessible from other exchanges, but circuits are located only in PSEG facilities. It is considered the primary backup for the NETS system.

NRC (ENS)

The Emergency Notification System (ENS) is a dedicated communications system with the NRC, which is part of the Federal Telecommunications System (FTS) and consists of direct lines to the NRC. FTS lines are used to provide general accident information. These telephones are installed in the Control Room, TSC, and the EOF.

This EAL is the cold condition equivalent of the hot condition EAL SU7.1.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, CU5 Example EAL #1, #2, #3 2. PSEG Nuclear Emergency Plan, Section 7 Communications 3. UFSAR 9.5.2 Communications Systems



EAL#: CU5.1




EAL#: CA6.1

EAL Category: C – Cold Shutdown / Refueling System Malfunction

EAL Subcategory: 6 – Hazardous Event Affecting Safety Systems

Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode

OPCON Applicability: 4 - Cold Shutdown, 5 – Refueling


EAL:

The occurrence of ANY Table C-5 hazardous event AND EITHER: Event damage has caused indications of degraded performance in at least one train

of a SAFETY SYSTEM needed for the current operating mode The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or

structure needed for the current operating mode

Table C-5 Hazardous Events

Seismic event (earthquake) Internal or external FLOODING event High winds or tornado strike FIRE EXPLOSION Other events with similar hazard characteristics

as determined by the Shift Manager



EAL#: CA6.1

Basis: Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not visible damage and does not meet the intent of this EAL. This EAL is intended to consider the collateral damage that has occurred due to the equipment failure. This EAL addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of safety of the plant. The first condition addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. The second condition addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. Escalation of the emergency classification level would be via IC CS1 or RS1.


Definitions:

EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.


FLOODING: A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:



EAL#: CA6.1


result in potential offsite exposures. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not visible damage.

This EAL is the cold condition equivalent of the hot condition EAL SA8.1.

EAL Bases Reference(s): 1. NEI 99-01, Rev. 06, CA6 Example EAL #1 2. HC.OP-AB.MISC-001(Q) Acts of Nature 3. HC.FP-PS.KC-0000 Fire and Medical Emergency Response Manual



EAL#: CA6.1




Attachment 1- Use of FPB Table

Attachment 1 – Use of Fission Product Barrier Table

OPCON Applicability: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown A point system is used to determine the Emergency Classification Level based on the Fission Product Barrier Table. Each Fission Product Barrier Loss and Potential Loss threshold is assigned a point value as noted below.

Perform the following:

1. Review all columns of the Fission Product Barrier Table and identify which need further review.

2. For each of the three barriers, determine the EAL with the highest point value. No more than one EAL should be selected for each barrier.

3. Add the point values for the three barriers.

4. Classify based on the point value sum as follows:

If the sum is: Classify as: EAL ECG

Att#

4, 5 ALERT ANY loss or ANY potential loss of either Fuel Clad or RCS

2

6 - 11 SITE AREA EMERGENCY

Loss or potential loss of ANY two barriers

OR Potential loss of 2 barriers with the loss of the 3rd barrier

3

12, 13 GENERAL EMERGENCY

Loss of ANY two barriers AND

Loss or potential loss of third barrier

4

5. Implement the appropriate ECG Attachment per above table.

6. Continue to review the Fission Product Barrier Table for changes that could result in emergency escalation or de-escalation.



Attachment 2 - Figures / Drawings

Attachment 2: EAL Bases Figures / Drawings

Figures/drawings referenced in the bases discussions of the EALs are listed in this Attachment.

Title Page No.

AC Power Distribution 2

RPV Level Instrument Ranges 5

Short-Term Containment Pressure Response Following Recirculation Line Break (4031 MWt)

6

Long-Term Containment Pressure Response Following Recirculation Line Break

7




AC Power Distribution

Hope Creek normally has three physically separate, independent 500 KV transmission lines connecting the Hope Creek 500 KV Switchyard with the Offsite Power Distribution Network (PJM). The three sources are:

• 500 KV Hope Creek - Salem Crosstie line.

• The Red Lion Line (referred to as the 5015 line) is a 30.1 mile tie to the Red LionSwitching Station located at Delaware City next to the river and feeds the 500 KVSwitchyard Bus Section 3.

• The New Freedom Line (referred to as the 5023 line) is a 42.9 mile tie to the NewFreedom Switching Station located northeast of Hope Creek in Camden County andfeeds the 500 KV Switchyard Bus Section 5.

Power is distributed from the 500 KV switchyard to a 13.8 KV ring bus. Station electrical loads are supplied from the 13.8 KV ring bus through two physically independent auxiliary power systems via Station Service Transformers which supply vital and non-vital station loads. Station Service Transformers 1AX501 and 1BX501 normally supply the 4.16 KV vital buses. The four 4.16 KV vital buses can be supplied by either 1AX501 or 1BX501. Two of the four vital buses are normally provided power from 1AX501 with alternate power from 1BX501; the other two are normally supplied power from 1BX501 with alternate power from 1AX501. Loss of the normal power supply to a 4.16 KV vital bus initiates a fast transfer (alternate feeder breaker closes) to the alternate source, provided power is available.

Additionally, each 4.16 KV vital bus has an Emergency Diesel Generator (EDG) which will automatically start and provide power to the bus in the event of a sustained loss of power to its associated vital bus. Additional automatic EDG starts are initiated on degraded power conditions on both 1AX501 and 1BX501, or under LOCA conditions. EDGs will not automatically provide power to the bus unless the bus has a sustained loss of power.




AC Power Distribution (cont’d)




RPV Level Instrument Ranges




Short-Term Containment Pressure Response Following Recirculation Line Break (4031 MWt)




Long-Term Containment Pressure Response Following Recirculation Line Break



Attachment 3 – Definitions

Attachment 3 – Definitions Selected words in the ECG Initiating Conditions (ICs) and Emergency Action Levels (EALs) have been set in all capital letters and bolded.These words are defined terms having specific meanings as they relate to this document and the definitions of these terms are provided below and in the basis for the EAL that the word is used in.

ALERT: Events are in process, or have occurred, which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of hostile action. Any releases are expected to be small fractions of the EPA Protective Action Guideline exposure levels.

CONFINEMENT BOUNDARY: Is the barrier(s) between areas containing radioactive substances and the environment and includes the multi-purpose canister (MPC) and, for the purposes of this EAL, the associated cask shielding.

CONTAINMENT CLOSURE: Is the procedurally defined action taken to secure the Containment (Primary or Secondary) and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

EMERGENCY ACTION LEVEL (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level. EMERGENCY CLASSIFICATION LEVEL: One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are: Unusual Event (UE), Alert, Site Area Emergency (SAE) and General Emergency (GE). EPA PAGs: Environment Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem CDE Thyroid. Actual or projected offsite exposures in excess of the EPA PAGs requires HCGS to recommend protective actions for the general public to offsite planning agencies. EXPLOSION: A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and evidence of heat are observed.




FISSION PRODUCT BARRIER THRESHOLD: A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. FLOODING: A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area. GENERAL EMERGENCY (GE): Events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTIONS that result in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.


HOSTILE ACTION: An act toward Salem or Hope Creek or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate PSEG to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILES, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile Action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on Salem or Hope Creek. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the OCA).

HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

IDENTIFIED LEAKAGE: As defined in T/S, shall be leakage into collection systems, such as pump seals or valve packing leaks, that is captured and conducted to a sump or collecting tank, or, shall be leakage into the containment atmosphere from sources that are both specifically located and known either not to interface with the operation of the leakage detection system or not to be PRESSURE BOUNDRY LEAKAGE.


IMPEDE(D): Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed). INITIATING CONDITION: An event or condition that aligns with the definition of one of the four EMERGENCY CLASSIFICATION LEVELS by virtue of the potential or actual effects or consequences.

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.






PRESSURE BOUNDARY LEAKAGE: As defined in T/S, shall be leakage through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROJECTILE: An object directed toward a Nuclear Power Plant that could cause concern for its continued operability, reliability, or personnel safety.


RCS INTACT: The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). REFUELING PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal comprise the refueling pathway. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2): Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could result

in potential offsite exposures.

SECURITY CONDITION: Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION.

SITE AREA EMERGENCY (SAE): Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or hostile actions that result in intentional damage or malicious acts; (1) toward site personnel or equipment that could lead to the likely failure of or; (2) that prevent effective access to




equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guidelines exposure levels beyond the site boundary.

UNIDENTIFIED LEAKAGE: As defined in T/S, shall be all leakage which is not IDENTIFIED LEAKAGE.



UNUSUAL EVENT (UE): Events are in process or have occurred which indicate a potential degradation in the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.



VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not visible damage.



Attachment 4 - Abbreviations & Acronyms

Attachment 4 – Glossary of Abbreviations & Acronyms Acronyms and Abbreviations used in the ECG and ECG basis document are listed in this attachment.

AAAG - Accident Assessment Advisory Group (Delaware) AC - Alternating Current ADMSS - Administrative Support Supervisor - TSC ADS - Automatic Depressurization System AH - Amp-Hour ALARA - As Low As Reasonably Achievable APRM - Average Power Range Monitor ARI - Alternate Rod Insertion ARM - Area Radiation Monitor ASAP - As Soon As Possible ASM - Administrative Support Manager ATWS - Anticipated Transient Without Scram BKGD - Background BKR - Breaker (electrical circuit) BLDG - Building BNE - Bureau of Nuclear Engineering (NJDEPE) BWR - Boiling Water Reactor CACS - Containment Atmosphere Control System CAS - Central Alarm Station CCPM - Corrected Counts per Minute CEDE - Committed Effective Dose Equivalent CDE - Committed Dose Equivalent CFR - Code of Federal Regulations CIS - Containment Isolation System CNTMT/CT - Containment (Barrier) CoC - Certificate of Compliance CO2 - Carbon Dioxide CP - Control Point CPM - Counts Per Minute CPS - Counts Per Second CR - Control Room CRS - Control Room Supervisor CREF - Control Room Emergency Filter System CRIDS - Control Room Integrated Display System CRD - Control Rod Drive CSS - Core Spray System




DC - Direct Current DAPA - Drywell Atmosphere Post Accident (Radiation monitor) DDE - Deep Dose Equivalent DEI - Dose Equivalent Iodine DEMA - Delaware Emergency Management Agency DEP - Department of Environmental Protection (NJ) DHS - Department of Homeland Security DID - Direct Inward Dial (phone system) DLD - Drywell Leak Detection DOE - Department of Energy DOT - Department of Transportation DPCC/DCR - Discharge Prevention, Containment, & Countermeasures/Discharge

Cleanup & Removal Plan DPM - Disintegrations per Minute DRCF - Dose Rate Conversion Factor EACS - ESF Equipment Area Cooling System EAL - Emergency Action Level EAS - Emergency Alert System (Broadcast) EC - Emergency Coordinator ECCS - Emergency Core Cooling Systems ECG - Event Classification Guide ECL - Emergency Classification Level EDG - Emergency Diesel Generator EDO - Emergency Duty Officer EERC - Energy & Environmental Resource Center (Old NTC) EMRAD - Emergency Radio (NJ) ENC - Emergency News Center ENS - Emergency Notification System (NRC) EOC - Emergency Operations Center (NJ & DE) EOF - Emergency Operations Facility EOP - Emergency Operating Procedure EPA - Emergency Preparedness Advisor EPA - Environmental Protection Agency EPC - Emergency Preparedness Coordinator EPG - Emergency Procedure Guideline EPIP/EPEP - Emergency Plan Implementing Procedure EPZ - Emergency Planning Zone EQPT - Equipment ERDS - Emergency Response Data System ERM - Emergency Response Manager ERO - Emergency Response Organization ESF - Engineered Safety Feature ESSX - Electronic Switch System Exchange (Centrex)




FAA - Federal Aviation Administration FBI - Federal Bureau of Investigation FC - Fuel Clad (Barrier) FEMA - Federal Emergency Management Agency FFD - Fitness For Duty FRVS - Filtration, Recirculation and Ventilation System FSAR - Final Safety Analysis Report FTS - Federal Telecommunications System (NRC) GE - General Emergency HCGS - Hope Creek Generating Station HCTL - Heat Capacity Temperature Limit HEPA - High Efficiency Particulate Absorbers HOO - Headquarters Operations Officer HPCI - High Pressure Coolant Injection HTV - Hardened Torus Vent HVAC - Heating, Ventilation & Air Conditioning HWCI - Hydrogen Water Chemical Injection HX - Heat Exchanger IAW - In Accordance With IC - Initiating Condition ICMF - Initial Contact Message Form IDLH - Immediately Dangerous to Life and Health IPEEE - Individual Plant Examination of External Events IRM - Intermediate Range Monitor ISFSI - Independent Spent Fuel Storage Installation I/S - In Service Keff - Effective Neutron Multiplication Factor KI - Potassium Iodide KV - Kilovolt LAC - Lower Alloways Creek LCO - Limiting Condition for Operation LDC - Learning Development Center (aka – NAB or TB2) LDE - Lens Dose Equivalent LEL - Lower Explosive Limit LER - Licensing Event Report LFL - Lower Flammability Limit LLD - Lowest Level Detectable LOCA - Loss of Coolant Accident LOP - Loss of Offsite Power




LPCI - Low Pressure Coolant Injection LPZ - Low Population Zone LWR - Light Water Reactor MCR - Main Control Room MDA - Minimum Detectable Amount MEA - Minimum Exclusion Area MEES - Major Equipment & Electrical Status (Form) MET - Meteorological MOU - Memorandum of Understanding MRO - Medical Review Officer MSCRWL - Minimum Steam Cooling RPV Water Level MSCP - Minimum Steam Cooling Pressure MSIV - Main Steam Isolation Valve MSIVSS - Main Steam Isolation Valve Sealing System MSL - Main Steam Line MSL - Mean Sea Level mR - milliRoentgen MW - Megawatt NAB - Nuclear Administrative Building NAWAS - National Attack Warning Alert System NCO - Nuclear Control Operator NDAB - Nuclear Department Administration Building (TB2) NEI - Nuclear Energy Institute NEO - Nuclear Equipment Operator NESP - National Environmental Studies Project NETS - Nuclear Emergency Telecommunications System NFE - Nuclear Fuels Engineer NFPB - Normal Full Power Background NJSP - New Jersey State Police NOAA - National Oceanographic and Atmospheric Administration NORAD - North American Aerospace Defense Command NPP - Nuclear Power Plant NPV - North Plant Vent NRC - Nuclear Regulatory Commission NSSSS - Nuclear Steam Supply Shutoff System NUMARC - Nuclear Management and Resources Council NWS - National Weather Service OBE - Operating Basis Earthquake OCA - Owner Controlled Area ODCM - Offsite Dose Calculation Manual OEM - Office of Emergency Management (NJ) OHA - Overhead Annunciator




OPCON - Operational Condition ORO - Offsite Response Organization OSB - Operational Status Board (Form) OSC - Operations Support Center PA - Protected Area PA - Public Address PAG - Protective Action Guideline PAR - Protective Action Recommendation PC - Primary Containment (Barrier) PCIG - Primary Containment Instrument Gas System PCIS - Primary Containment Isolation System PPC - Plant Process Computer PRM - Process Radiation Monitor PSEG - Public Service Enterprise Group PSIG - Pounds Square Inch Gauge R - Roentgen RAD - Radiation RAL - Reportable Action Level RC - Reactor Coolant RCA - Radiologically Controlled Area RCAM - Repair and Corrective Action Mission RCIC - Reactor Core Isolation Cooling RCS - Reactor Coolant System (Barrier) REM - Roentgen Equivalent Man RHR - Residual Heat Removal (Containment Heat Removal) RM - Recovery Manager RM - Radiation Monitor RMO - Recovery Management Organization RMS - Radiation Monitoring System RPS - Radiation Protection Supervisor RPS - Reactor Protection System RPV - Reactor Pressure Vessel RRCS - Redundant Reactivity Control System RSM - Radiological Support Manager RWCU - Reactor Water Cleanup (System) SACS - Safety Auxiliaries Cooling System SAE - Site Area Emergency SAM - Severe Accident Management SAS - Secondary Alarm Station (Security) SBO - Station Blackout SCBA - Self Contained Breathing Apparatus SCP - Security Contingency Procedure




SDE - Shallow Dose Equivalent SDM - Shutdown Margin SLC - Standby Liquid Control SJAE - Steam Jet Air Ejector SM - Shift Manager SNM - Special Nuclear Material SNSS - Senior Nuclear Shift Supervisor SOS - Systems Operations Supervisor (Security) SPDS - Safety Parameter Display System SPV - South Plant Vent SRM - Source Range Monitor SRO - Senior Reactor Operator SRPT - Shift Radiation Protection Technician SRV - Safety Relief Valve SSC - Structure, System or Component SSCL - Station Status Checklist SSE - Safe Shutdown Earthquake SSWS - Station Service Water System SSNM - Strategic Special Nuclear Material STA - Shift Technical Advisor SW - Service Water TAF - Top of Active Fuel TDR - Technical Document Room TEDE - Total Effective Dose Equivalent TIP - Traversing Incore Probe TLV - Threshold Limit Value T/S - Technical Specifications TSC - Technical Support Center TSS - Technical Support Supervisor TSTL - Technical Support Team Leader TSTM - Technical Support Team Member UE - Unusual Event UFSAR - Updated Final Safety Analysis Report UHS - Ultimate Heat Sink USCG - United States Coast Guard VDC - Volts Direct Current WB - Whole Body



Attachment 5 – EAL Cross-Reference

Attachment 5 – HCGS-to-NEI 99-01 EAL Cross-Reference This cross-reference is provided to facilitate association and location of a Hope Creek Generating Station EAL within the NEI 99-01, Rev. 06, IC/EAL identification scheme. Further information regarding the development of the HCGS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

HCGS NEI 99-01

EAL IC Example EAL

RU1.1 AU1 1, 2

RU1.2 AU1 3

RU2.1 AU2 1

RU4.1 E-HU1 1

RA1.1 AA1 1

RA1.2 AA1 2

RA1.3 AA1 3

RA1.4 AA1 4

RA2.1 AA2 1

RA2.2 AA2 2

RA2.3 AA2 3

RA3.1 AA3 1

RA3.2 AA3 2

RS1.1 AS1 1

RS1.2 AS1 2

RS1.3 AS1 3

RS2.1 AS2 1

RG1.1 AG1 1

RG1.2 AG1 2

RG1.3 AG1 3

RG2.1 AG2 1

CU1.1 CU1 1

CU1.2 CU1 2




HCGS NEI 99-01

EAL IC Example EAL

CU2.1 CU2 1

CU3.1 CU3 1

CU3.2 CU3 2

CU4.1 CU4 1

CU5.1 CU5 1, 2, 3

CA1.1 CA1 1

CA1.2 CA1 2

CA2.1 CA2 1

CA3.1 CA3 1, 2

CA6.1 CA6 1

CS1.1 CS1 1, 2

CS1.2 CS1 3

CG1.1 CG1 1

CG1.2 CG1 2

HU1.1 HU1 1, 2, 3

HU2.1 HU2 1

HU3.1 HU3 1

HU3.2 HU3 2

HU3.3 HU3 3

HU3.4 HU3 4

HU4.1 HU4 1

HU4.2 HU4 2

HU4.3 HU4 3

HU4.4 HU4 4

HU7.1 HU7 1

HA1.1 HA1 1, 2

HA5.1 HA5 1

HA6.1 HA6 1




HCGS NEI 99-01

EAL IC Example EAL

HA7.1 HA7 1

HS1.1 HS1 1

HS6.1 HS 6

HS7.1 HS7 1

HG7.1 HG7 1

SU1.1 SU1 1

SU3.1 SU2 1

SU4.1 SU3 1

SU4.2 SU3 2

SU5.1 SU4 1, 2, 3

SU6.1 SU5 1, 2

SU7.1 SU6 1, 2, 3

SA1.1 SA1 1

SA3.1 SA2 1

SA6.1 SA6 1

SA8.1 SA9 1

SS1.1 SS1 1

SS2.1 SS8 1

SS6.1 SS5 1

SG1.1 SG1 1

SG2.1 SG8 1




Fission Product Barrier EALs

HCGS NEI 99-01 Threshold Barrier Threshold

FB1.P FC P.Loss 2.A

FB2.P FC P.Loss 6.A

FB1.L FC Loss 2.A

FB2.L FC Loss 4.A

FB3.L FC Loss 1.A

FB4.L FC Loss 6.A

RB1.P RCS P.Loss 3.A

RB2.P RCS P.Loss 6.A

RB1.L RCS Loss 2.A

RB2.L RCS Loss 1.A

RB3.L RCS Loss 3.A

RB4.L RCS Loss 3.B

RB5.L RCS Loss 6.A

CB1.P CMT P.Loss 2.A


CB3.P CMT P.Loss 1.B

CB4.P CMT P.Loss 1.C



CB1.L CMT Loss 1.A

CB2.L CMT Loss 1.B

CB3.L CMT Loss 3.A

CB4.L CMT Loss 3.B

CB5.L CMT Loss 3.C

CB6.L CMT Loss 6.A

HCGS ECG – EAL Technical Basis EP-HC-325-236


Attachment 6 - Rad Setpoint Calculation

Attachment 6 – Hope Creek EAL Rad Set-Point Calculation Document

Purpose:

This is a reference document that contains the methodology and calculations used in developing the thresholds for radiological release based Emergency Action Levels (EALs). The radiological EALs covered under this document are based on EALs AU1, AA1, AS1 and AG1 in NEI-99-01, Rev. 06, “Development of Emergency Action Levels for Non-Passive Reactors.”

Reference Materials:

• NEI 99-01, Rev. 06 - Development of Emergency Action Levels for Non-Passive Reactors

• Hope Creek ODCM Rev. 27

• EPA 400-R-92-001, Manual for Protective Action Guides and Protective Actions for NuclearIncidents

Terms & Calculation Constants and origin:

• ODCM – Offsite Dose Calculation Manual

• Hours in one year: 365.25 days X 24 hrs/day = 8766 hours

• EDE – Effective Dose Equivalent

• CDE - Committed Dose Equivalent

• CEDE - Committed Effective Dose Equivalent = CDE X Weighting Factor (thyroid per 10 CFR 20)

• TEDE – Total Effective Dose Equivalent = EDE + CEDE

• FRVS – Filtration Recirculation Vent System

• NPV – North Plant Vent

• SPV – South Plant Vent

• HTV – Hardened Torus Vent

• PAG – Protective Action Guideline: Per EPA = 1000mRem TEDE dose or 5000 mRem thyroiddose. Actual or projected values above these guidelines will require offsite protective actions tobe implemented.

• ODCM Rad Effluent Limit - 500 mRem/year is a total site Noble Gas limit that includes Salem 1,Salem 2 and Hope Creek. Therefore, Hope Creek will have an administratively controlled limit of½ the total site limit or 250 mRem/year for Unusual Event EAL calculation purposes.

• Allocation Factor (AF) = .5 – As defined in the Hope Creek ODCM, (page 85) this is anadministrative control imposed to ensure that the combined releases from Salem Units 1 and 2




and Hope Creek will not exceed the regulatory limit from the site. The Site AF is only used in the UE EALs.

• X/Q = Site Specific Atmospheric dispersion to the site boundary.

Hope Creek Value = 2.14E-06 sec/m3. Origin – Hope Creek ODCM, Rev. 27, Table 2-2, Parameters for Gaseous Alarm Setpoint Determinations.

• DRCF = Site Specific Dose Rate Conversion Factor. Hope Creek = 7.8E+03 mrem/year per uCi/ m3. Origin – Hope Creek ODCM, Rev. 24, Table C-1, Effective Dose Factors, Noble Gases – Total Body and Skin – Total Body Effective Dose Factor.

Index: (Radiological Release EAL Calculations) Hope Creek EALS: (NEI 99-01, revision 6 equivalent EAL) Page #: Unusual Event EAL RU1.1 ((AU1(1)) – (Default Release Rate EAL) 4 Unusual Event EAL RU1.2 ((AU1(2)) – (Liquid Rad Alarm) No Calc needed - Unusual Event EAL RU1.3 ((AU1(3)) – (Gaseous Sample Analysis) 5 Unusual Event EAL RU1.3 ((AU1(3)) – (Liquid Sample Analysis) No Calc needed - Alert EAL RA1.1 ((AA1(1)) – (Default Release Rate EAL) 7 Alert EAL RA1.2 ((AA1(2)) – (Dose Assessment) 8 Alert EAL RA1.3 ((AA1(3)) – (Liquid Sample Analysis) No Calc needed - Alert EAL RA1.4 ((AA1(4)) – (Field Survey Dose Rate) 9 Alert EAL RA1.4 ((AA1(4)) – (Field Survey Iodine) 10 Site Area Emergency EAL RS1.1 ((AS1(1)) – (Default Release Rate EAL) 12 Site Area Emergency EAL RS1.2 ((AS1(2)) – (Dose Assessment) 13 Site Area Emergency EAL RS1.3 ((AS1(3)) – (Field Survey Dose Rate) 14 Site Area Emergency EAL RS1.3 ((AS1(3)) – (Field Survey Iodine) 15 General Emergency EAL RG1.1 ((AG1(1)) – (Default Release Rate EAL) 17 General Emergency EAL RG1.2 ((AG1(2)) – (Dose Assessment) 18 General Emergency EAL RG1.3 ((AG1(3)) – (Field Survey Dose Rate) 19 General Emergency EAL RG1.3 ((AG1(3)) – (Field Survey Iodine) 20




Calculation for: Unusual Event EAL RU1.1 – (Default Release Rate EAL) Objective of Calculation:

Provide a Hope Creek Radiological Release Rate value that equates to a Release that is >2 times the ODCM limit of 500 mRem/year.

Discussion: The ODCM limit of 500 mRem/year is a total site limit that includes Salem 1, Salem 2 and Hope Creek. Therefore, Hope Creek will have an administratively controlled limit of ½ the total site limit or 250 mRem/year for EAL calculation purposes. This EAL does not include Iodine Release Rates, since the Hope Creek Effluent monitors (FRVS, NPV, SPV and HTV) do not have Iodine detectors. Release Rate = Total Noble Gas Release Rate from Hope Creek (FRVS, NPV, SPV, HTV) which would result in a TEDE Dose Rate of 250 mRem/year. The EAL value will be 2 times the release rate. Derivation / Calculation: ODCM Limit Calculation for Noble Gas:

Release Rate (uCi/Sec) = ODCM Limit �mRem

year �∗(Site Allocation Factor)

�ODCMXQ�∗(ODCM DRCF)

ODCM Limit = 500 mRem/Year Hope Creek ODCM X/Q = 2.14E-06 sec/m3 Hope Creek ODCM DRCF = 7.80E+03 mRem/yr/uCi/m3 Site Allocation Factor = 5.00E-01

Release Rate (uCi/Sec) = )///Re0380.7(*)sec/0614.2(

)0100.5(*)/Re500(33 mCiyrmmEmE

Eyrmmm+−

−

Release Rate = 1.50E+04 uCi/Sec (Also the Tech Spec/ODCM Limit Release Rate Value) EAL Value = 2 times the Release Rate UE EAL Value: (EAL # RU1.1)

Total Hope Creek Noble Gas Release Rate >3.00E+04 mCi/sec




Calculation for: Unusual Event EAL RU1.3 – (Sample Analysis Concentration) Objective of Calculation:

Provide a Radiological Release Noble Gas and Iodine Sample Concentration that equates to a Release that is >2 times the ODCM limit of 500 mRem/year.

Discussion: The ODCM limit of 500 mRem/year (Noble Gas/Total Body) and 1500mRem/year (I-131/Child Thyroid) is a total site limit that includes Salem 1, Salem 2 and Hope Creek. Therefore, Hope Creek will have an administratively controlled limit (allocation factor) of ½ the total site limit or 250 mRem/year (Noble Gas/Total Body) and 750 mRem/year (I-131/Child Thyroid) for EAL calculation purposes. This allocation factor is used in the calculation that derived the NG and Iodine release rates. Hope Creek has three release points for which sample concentration thresholds are needed.

• FRVS – Filtration Recirculation Vent System • NPV – North Plant Vent • SPV – South Plant Vent

Derivation / Calculation: Calculation of the threshold sample concentrations are as follows:

Formula: Concentration (uCi/cc) = teVentFlowRaFactorConversion

leaseRateODCM*

2*Re

FRVS Noble Gas Sample Concentration = cfmx

CiE9000472

2sec*/0450.1 m+ = 7.10E-03 mCi/cc

FRVS I-131 Sample Concentration = cfmx

CiE9000472

2sec*/0174.1 m+ = 8.20E-06 mCi/cc

NPV Noble Gas Sample Concentration= cfmEx

CiE0419.4472

2sec*/0450.1+

+ m = 1.52E-03 mCi/cc

NPV I-131 Sample Concentration = cfmEx

CiE0419.4472

2sec*/0174.1+

+ m = 1.80E-06 mCi/cc

SPV Noble Gas Sample Concentration = cfmEx

CiE0540.4472

2sec*/0450.1+

+ m = 1.44E-04 mCi/cc




SPV I-131 Sample Concentration = cfmEx

CiE0540.4472

2sec*/0174.1+

+ m = 1.68E-07 mCi/cc

Where:

• ODCM limit Release Rate of 1.50E+04 µCi/sec per calculation performed for EAL AU1.1.

• ODCM limit (Thyroid/I-131) Release Rate of 17.4 uCi/Sec as per ODCM, Rev. 27, Section 2.3.2 -

• 2 = EAL criteria of 2X Tech Spec/ODCM value • 472 = conversion factor (28,317 cc/ft3 x 1 min./60 sec.) • 9000 cfm = FRVS Vent Flow (maximum) (ODCM Table 2-2) • 4.19E+04 cfm = NPV Vent Flow (maximum) (ODCM Table 2-2) • 4.40E+05 cfm = SPV Vent Flow (maximum) (ODCM Table 2-2)

UE EAL Values: (EAL# RU1.3)

FRVS Noble Gas Sample Concentration > 7.10E-03 mCi/cc FRVS I-131 Sample Concentration > 8.20E-06 mCi/cc NPV Noble Gas Sample Concentration > 1.52E-03 mCi/cc NPV I-131 Sample Concentration > 1.80E-06 mCi/cc SPV Noble Gas Sample Concentration > 1.44E-04 mCi/cc SPV I-131 Sample Concentration > 1.68E-07 mCi/cc




Calculation for: ALERT EAL RA1.1 – (Default Release Rate EAL) Objective of Calculation:

Provide a Radiological Release of gaseous radioactivity resulting in offsite dose greater than 10 mRem TEDE or 50 mRem thyroid CDE.

Discussion: This EAL does not include Iodine Release Rates, since the Hope Creek Effluent monitors (FRVS, NPV, SPV and HTV) do not have Iodine detectors. Release Rate = Total Noble Gas Release Rate from Hope Creek (FRVS, NPV, SPV, HTV) which would result in a TEDE Dose Rate of 10 mRem. Derivation / Calculation: ODCM Limit Calculation for Noble Gas: Release Rate (uCi/Sec) = (1% of PAG) mRem (accumulated in 1 hour) (ODCM X/Q)*(ODCM DRCF) 1% of PAG = 1% of 1000 mRem or 10 mRem (dose accumulated in 1 hour) Hope Creek ODCM X/Q = 2.14 E-06 sec/m3 HC ODCM DRCF = 8.9 E-01 mRem/hr/uCi/m3 (7.80E+03 mRem/yr/uCi/m3/8766hrs/yr) Site Allocation Factor = not used for Alert, SAE and GE EALs Release Rate Noble Gas (uCi/Sec) = 10 mRem (dose accumulated in 1 hour) = 5.25 E+06 uCi/sec (2.14E-06 sec/m3)*( 8.9E-01 mRem/hr/uCi/m3) ALERT EAL Value: (EAL# RA1.1)

Total Hope Creek Noble Gas Release Rate >5.25E+06 mCi/sec




Calculation for: ALERT EAL RA1.2 – (Dose Assessment)

Objective of Calculation:

Using actual meteorology, provide a Station Status Checklist (SSCL) dose assessment threshold TEDE 4-Day Dose value that is equivalent to a TEDE dose of >10 mRem and a Thyroid-CDE Dose of >50 mRem.

Discussion: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 1% of the EPA Protective Action Guides (PAGs). Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits. Derivation / Calculation: The dose assessment output on the SSCL is reported at varying distances from the plant as a TEDE 4-Day dose. Exceeding the EAL value at a distance at or beyond the Minimum Exclusion Area (MEA) satisfies this ALERT EAL threshold. Dose Assessment using actual meteorological data provides an accurate indication of release magnitude. The use of dose assessment based EALs is therefore preferred over the use of Default Release Rate based EALs which utilize calculations which have built-in inaccuracies because ODCM default Meteorological data is used.

ALERT EAL Values: (EAL# RA1.2) Dose Assessment TEDE Dose >10 mRem Dose Assessment CDE Dose >50 mRem – CDE Dose values are based on Dose Assessment as input to MIDAS and NOT based on a default Noble Gas to Iodine Ratio




Calculation for: ALERT - EAL RA1.4 – (Field Survey Dose Rate)


Provide a Field Survey dose rate that equates to an offsite dose of >10 mRem/hr - closed window. Discussion: This EAL addresses radioactivity releases that result in field survey results (closed window) dose rates greater than 10 mR/hr expected to continue for 60 minutes or longer at or beyond the site boundary. This value exceeds 1% of the EPA Protective Action Guides (PAGs). Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant.

Derivation / Calculation: A Field Measured (closed window) Dose Rate of >10 mRem/hr corresponds directly to a dose values that exceed 1% of the EPA Protective Actions Guides (PAGs). ALERT EAL Value: (EAL# RA1.4) Closed Window Dose Rate >10 mRem/hr




Calculation for: ALERT - EAL RA1.4 – (Field Survey Iodine) Objective of Calculation:

Provide a Field Survey Sample Analysis value that equates to an offsite release that would result in a dose of >50 mRem Thyroid CDE at or beyond the PROTECTED AREA Boundary for one hour of inhalation. Discussion: This EAL addresses a radioactivity release field survey I-131 sample concentration or count rate that would result in a Thyroid CDE dose of greater than 50 mRem for one hour of inhalation at or beyond the site boundary. This value exceeds 1% of the EPA Protective Action Guides (PAGs). Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant.

The Iodine-131 field survey sample concentration and count rate threshold is based on I-131 dose conversion factors (DCFs) from EPA-400. The thresholds are based on a Thyroid-CDE Dose Rate of >50 mRem/hr for I-131. Field Survey I-131 Sample Analysis results are provided as both a sample concentration in units of uCi/cc for field samples counted in a Multi-Channel-Analyzer (MCA) and as a count rate reading in units of Corrected Counts per Minute (CCPM) for field samples analysis obtained using a radiation count rate meter such as a RM-14 or E-140N with a HP260 probe attached.

Derivation / Calculation: The release sample concentration calculations are as follows. The sample concentration is calculated using the I-131 Dose Conversion Factor from EPA-400: Solving the following equation for µCi/cc: mRem/hr = (µCi/cc)(Dose Conversion Factor) Then; I-131 Sample Concentration (µCi/cc) = ( 50 mRem/Hr ) = 3.85E-08 mCi/cc 1.30E+09 m/Rem/uCi/cc/hr Where 1.30E+09 mRem/µCi/cc/hr is the Dose Conversion Factor from EPA-400, Table 5-4, Thyroid Dose, and includes the EPA breathing rate. The Corrected Counts per Minute reading is calculated using the I-131 Sample concentration, and factors for using an RM-14 or E-140N with an HP260 probe. Solving the following equation for CCPM:




µCi/cc = CCPM_______________________________________________________ (Detector Efficiency)(Collection Efficiency)(Conversion Factor - DPM to µCi)(Volume - ft3 )(Conversion Factor - cc to ft3 ) Then; CCPM = (3.85E-08 mCi/cc) (2.00E-03 CCPM/DPM) (0.9) (2.22E+06 DPM/mCi) * (10 ft3) (2.832E+04cc/ft3) = 4.36E+01 CCPM Rounded to: 45 CCPM (for ease of reading value on RM-14 or E-140 instrument) Where: CCPM = Corrected Counts per Minute using an RM-14 or E-140N with an HP260 probe. CCPM = Gross CPM – Bkg CPM 2.00E-03 = Detector Efficiency - CCPM/DPM 0.9 (or 90%) = Collection Efficiency 2.22E+06 = Conversion factor - DPM/µCi 10 ft3 = Volume 2.832E+04 = Conversion factor - cc to ft3 ALERT EAL Values: (EAL# RA1.4)

I-131 Concentration >3.85E-08 mCi/cc HP 260 Probe Reading Value >45 CCPM




Calculation for: SITE AREA EMERGENCY - EAL RS1.1 – (Default Release Rate EAL)


Provide a Radiological Release Rate value that equates to a Release resulting in an offsite dose of >100 mrem TEDE at or beyond the site boundary.

Discussion: This IC addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

The monitor reading EALs should be determined using a dose assessment method that back calculates from the dose values specified in the IC. Since doses are generally not monitored in real-time, it is suggested that a release duration of one hour be assumed, and that the EALs be based on a site specific boundary (or beyond) dose of 100 mrem whole body. Iodine Release Rates for this EAL are excluded since the Plant Vent Radiation Monitoring System does not include an Iodine detector. The meteorology and source term used are the same as used for determining AU1 and AA1 monitor reading EALs. Release Rate = Total Noble Gas Release Rate from Hope Creek which would result in a TEDE Dose Rate of >100mRem/hr at the site boundary or beyond. Derivation / Calculation: ODCM Limit Calculation for Noble Gas:

Release Rate (uCi/Sec) = )(*)/(

)1(Re)%10(DRCFODCMQXODCM

hourdinaccumulatemmofPAG

10% of PAG = 100 mRem/hr dose accumulated in 1 hour Hope Creek ODCM X/Q = 2.14E-06 sec/m3 HC ODCM DRCF = 8.9E-01 mRem/hr/uCi/m3 (7.80E+03 mRem/yr/uCi/ m3 / 8766 hrs/yr) Site Allocation Factor = not used for Alert, SAE and GE EALs

Release Rate (uCi/Sec) = )3///Re019.8(*)sec/0614.2(

)1(Re1003 muCihrmmEmE

hourlatedindoseaccumumm−−

SAE EAL Value: (EAL# RS1.1)

Total Hope Creek Noble Gas Release Rate >5.25E+07 uCi/Sec




Calculation for: SITE AREA EMERGENCY - EAL RS1.2 – (Dose Assessment)


Using actual meteorology, provide a dose assessment SSCL threshold TEDE 4-Day Dose value that is equivalent to a TEDE dose of >100 mRem and a Thyroid-CDE Dose of >500 mRem.

Discussion: This IC addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Derivation / Calculation: The dose assessment output on the SSCL is reported at varying distances from the plant as a TEDE 4-Day dose. Exceeding the EAL value at a distance at or beyond the Minimum Exclusion Area (MEA) satisfies this SAE EAL threshold. Dose Assessment using actual meteorological data provides an accurate indication of release magnitude. The use of dose assessment based EALs is therefore preferred over the use of Default Release Rate based EALs which utilize calculations which have built-in inaccuracies because ODCM default Meteorological data is used. SAE EALs Values: (EAL# RS1.2) Dose Assessment TEDE 4-Day Dose >1.0E+02 mRem Dose Assessment CDE Dose >5.00E+02 mRem - based on Dose Assessment as input to MIDAS and NOT based on a default Noble Gas to Iodine Ratio




Calculation for: SITE AREA EMERGENCY - EAL RS1.3 – (Field Survey Dose Rate) Objective of Calculation:

Provide a Field Survey dose rate that equates to an offsite dose of >100 mRem/hr - closed window. Discussion: This IC addresses radioactivity releases that result in field survey results (closed window) dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer at or beyond the site boundary. This value exceeds 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Derivation / Calculation: A Field Measured (closed window) Dose Rate of > 100 mRem/hr corresponds directly to a dose values that exceed 10% of the EPA Protective Actions Guides (PAGs). SAE EAL Value: (EAL# RS1.3) Closed Window Dose Rate >100 mRem/hr




Calculation for: SITE AREA EMERGENCY - EAL RS1.3 – (Field Survey Iodine) Objective of Calculation:

Provide a Field Survey Sample Analysis value that equates to an offsite release that would result in a dose of >500 mRem Thyroid CDE at or beyond the PROTECTED AREA Boundary. Discussion: This EAL addresses a radioactivity release field survey I-131 sample concentration or count rate that would result in a Thyroid CDE dose of greater than 500 mRem for one hour of inhalation at or beyond the site boundary. This value exceeds 10% of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

The Iodine-131 field survey sample concentration and count rate threshold is based on I-131 dose conversion factors (DCFs) from EPA-400. The thresholds are based on a Thyroid-CDE Dose Rate of >500 mRem/hr for I-131. Field Survey I-131 Sample Analysis results are provided as both a sample concentration in units of uCi/cc for field samples counted in a Multi-Channel-Analyzer (MCA) and as a count rate reading in units of Corrected Counts per Minute (CCPM) for field samples analysis obtained using a radiation count rate meter such as a RM-14 or E-140N with a HP260 probe attached.

Derivation / Calculation: The release sample concentration calculations are as follows. The sample concentration is calculated using the I-131 Dose Conversion Factor from EPA-400: Solving the following equation for µCi/cc: mRem/hr = (µCi/cc)(Dose Conversion Factor) Then;

I-131 Sample Concentration (µCi/cc) = ( 500130 09

m m hrE m m Ci cc hr

Re /. Re / / /+ m

) = 3.85E-07 mCi/cc

Where 1.30E+09 mRem/µCi/cc/hr is the Dose Conversion Factor from EPA-400, Table 5-4, Thyroid Dose, and includes the EPA breathing rate. The Corrected Counts per Minute reading is calculated using the I-131 Sample concentration, and factors for using an RM-14 or E-140N with an HP260 probe.




Solving the following equation for CCPM: µCi/cc = CCPM_______________________________________________________ (Detector Efficiency)(Collection Efficiency)(Conversion Factor - DPM to µCi)(Volume - ft3 )(Conversion Factor - cc to ft3 ) Then; CCPM = (3.85E-07 mCi/cc) (2.00E-03 CCPM/DPM) (0.9) (2.22E+06 DPM/mCi) * (10 ft3) (2.832E+04 cc/ft3) = 4.36E+02 CCPM Rounded to: 4.50E+02 CCPM (for ease of reading value on RM-14 or E-140 instrument) Where: CCPM = Corrected Counts per Minute using an RM-14 or E-140N with an HP260 probe. CCPM = Gross CPM – Bkg CPM 2.00E-03 = Detector Efficiency - CCPM/DPM 0.9 (or 90%) = Collection Efficiency 2.22E+06 = Conversion factor - DPM/µCi 10 ft3 = Volume 2.832E+04 = Conversion factor - cc to ft3 SAE EAL Values: (EAL# RS1.3)

I-131 Concentration >3.85E-07 mCi/cc HP 260 Probe Reading >4.50E+02 CCPM




Calculation for: GENERAL EMERGENCY - EAL RG1.1 – (Default Release Rate EAL) Objective of Calculation:

Provide a Radiological Release Rate value that equates to a Release resulting in an offsite dose of >1000 mRem TEDE at or beyond the site boundary.

Discussion: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude will require implementation of protective actions for the public.

The monitor reading EALs should be determined using a dose assessment method that back calculates from the dose values specified in the IC. Since doses are generally not monitored in real-time, it is suggested that a release duration of one hour be assumed, and that the EALs be based on a site specific boundary (or beyond) dose of >1000 mRem whole body. Iodine Release Rates for this EAL are excluded since the Plant Vent Radiation Monitoring System does not include an Iodine detector.

The meteorology and source term used are the same as used for determining RU1 and RA1 monitor reading EALs. Release Rate = Total Noble Gas Release Rate from Hope Creek which would result in a TEDE Dose Rate of >1000 mRem/hr at the site boundary or beyond. Derivation / Calculation: ODCM Limit Calculation for Noble Gas:

Release Rate (uCi/Sec) = )(*)/(

)1(Re)%100(DRCFODCMQXODCM

hourdinaccumulatemmofPAG

100% of PAG = 1000 mRem/hr dose accumulated in 1 hour Hope Creek ODCM X/Q = 2.14E-06 sec/m3 HC ODCM DRCF = 8.9E-01 mRem/hr/uCi/m3 (7.80E+03 mRem/yr/uCi/ m3 / 8766 hrs/yr) Site Allocation Factor = not used for SAE and GE EALs

Release Rate (uCi/Sec) = )3///Re019.8(*)sec/0614.2(

)1(Re10003 muCihrmmEmE

hourlatedindoseaccumumm−−

GE EAL Value: (EAL# RG1.1)

Total Hope Creek Noble Gas Release Rate >5.25E+08 uCi/Sec




Calculation for: GENERAL EMERGENCY - EAL RG1.2 – (Dose Assessment) Objective of Calculation:

Using actual meteorology, provide a dose assessment SSCL threshold TEDE 4-Day Dose value that is equivalent to a TEDE dose of >1000 mRem and a Thyroid-CDE Dose of >5000 mRem.

Discussion: This EAL addresses radioactivity releases that result in doses at or beyond the site boundary that exceed 100% of the EPA Protective Action Guides (PAGs). Releases of this magnitude will require implementation of protective actions for the public.

Derivation / Calculation: The dose assessment output on the SSCL is reported at varying distances from the plant as a TEDE 4-Day dose. Exceeding the EAL value at a distance at or beyond the Minimum Exclusion Area (MEA) satisfies this GE EAL threshold. Dose Assessment using actual meteorological data provides an accurate indication of release magnitude. The use of dose assessment based EALs is therefore preferred over the use of Default Release Rate based EALs which utilize calculations which have built-in inaccuracies because ODCM default Meteorological data is used. GE EALs Values: (EAL# RG1.2) Dose Assessment TEDE 4-Day Dose >1.0E+03 mRem Dose Assessment CDE Dose >5.00E+03 mRem - based on Dose Assessment as input to MIDAS and NOT based on a default Noble Gas to Iodine Ratio




Calculation for: GENERAL EMERGENCY - EAL RG1.3 – (Field Survey Dose Rate) Objective of Calculation:

Provide a Field Survey dose rate that equates to an offsite dose of >1000 mRem/hr - closed window. Discussion: This IC addresses radioactivity releases that result in field survey results (closed window) dose rates greater than 1000 mR/hr expected to continue for 60 minutes or longer at or beyond the site boundary. This value exceeds 100% of the EPA Protective Action Guides (PAGs). Releases of this magnitude will require implementation of protective actions for the public.

Derivation / Calculation: A Field Measured (closed window) Dose Rate of >1000 mRem/hr corresponds directly to a dose value that exceed the EPA Protective Actions Guides (PAGs). GE EAL Value: (EAL# RG1.3) Closed Window Dose Rate >1000 mRem/hr




Calculation for: GENERAL EMERGENCY - EAL RG1.3 – (Field Survey Iodine)


Provide a Field Survey Sample Analysis value that equates to an offsite release that would result in a dose of >5000 mRem Thyroid CDE at or beyond the PROTECTED AREA Boundary. Discussion: This EAL addresses a radioactivity release field survey I-131 sample concentration or count rate that would result in a Thyroid CDE dose of greater than 5000 mRem for one hour of inhalation at or beyond the site boundary. This value exceeds the EPA Protective Action Guides (PAGs). Releases of this magnitude will require implementation of protective actions for the public.

The Iodine-131 field survey sample concentration and count rate threshold is based on I-131 dose conversion factors (DCFs) from EPA-400. The thresholds are based on a Thyroid-CDE Dose Rate of >5000 mRem/hr for I-131. Field Survey I-131 Sample Analysis results are provided as both a sample concentration in units of uCi/cc for field samples counted in a Multi-Channel-Analyzer (MCA) and as a count rate reading in units of Corrected Counts per Minute (CCPM) for field samples analysis obtained using a radiation count rate meter such as a RM-14 or E-140N with a HP260 probe attached. Derivation / Calculation: The release sample concentration calculations are as follows. The sample concentration is calculated using the I-131 Dose Conversion Factor from EPA-400: Solving the following equation for µCi/cc: mRem/hr = (µCi/cc)(Dose Conversion Factor) Then;

I-131 Sample Concentration (µCi/cc) = (hrccCimmE

hrmm///Re0930.1

/Re5000m+

) = 3.85E-06 mCi/cc

Where 1.30E+09 mRem/µCi/cc/hr is the Dose Conversion Factor from EPA-400, Table 5-4, Thyroid Dose, and includes the EPA breathing rate. The Corrected Counts per Minute reading is calculated using the I-131 Sample concentration, and factors for using an RM-14 or E-140N with an HP260 probe. Solving the following equation for CCPM:




µCi/cc = CCPM_______________________________________________________ (Detector Efficiency)(Collection Efficiency)(Conversion Factor - DPM to µCi)(Volume - ft3 )(Conversion Factor - cc to ft3 ) Then; CCPM = (3.85E-06 mCi/cc) (2.00E-03 CCPM/DPM) (0.9) (2.22E+06 DPM/mCi) * (10 ft3) (2.832E+04 cc/ft3) = 4.36E+03 CCPM Rounded to: 4.50E+03 CCPM (for ease of reading value on RM-14 or E-140 instrument) Where: CCPM = Corrected Counts per Minute using an RM-14 or E-140N with an HP260 probe. CCPM – Gross CPM – BKG CPM 2.00E-03 = Detector Efficiency - CCPM/DPM 0.9 (or 90%) = Collection Efficiency 2.22E+06 = Conversion factor - DPM/µCi 10 ft3 = Volume 2.832E+04 = Conversion factor - cc to ft3 . GE EAL Values: (EAL# RG1.3)

I-131 Concentration >3.85E-06 mCi/cc HP 260 Probe Reading >4.50E+03 CCPM



Attachment 7 – Safe Operation & Shutdown Rooms/Areas Bases

Attachment 7 – Hope Creek Safe Operation & Shutdown Rooms/Areas Table R-3 & H-2 Bases

Background NEI 99-01 Revision 6 lCs AA3 and HA5 prescribe declaration of an Alert based on impeded access to rooms or areas due to radiation or hazardous gas concentrations where equipment necessary for normal plant operations, cooldown or shutdown is located. These areas are intended to be plant operating condition-Opcon or OC (mode) dependent. Specifically the Developers Notes for HA5 states:

The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area. The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

The list need not include the Control Room if adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas. Such features may include, but are not limited to, capability to draw air from multiple air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope. Site-Specific Analysis The Control Room is excluded in Table H-2 for all OCs since adequate engineered safety/design features are in place to preclude a Control Room evacuation due to the release of a hazardous gas at HCGS – a positive pressure is maintained in MCR during normal operation; the Recirculation mode of Control Room Emergency Filtration (CREF) does not maintain a positive pressure, but does filter and recirculate as well as condition (heat / cool) the control room air. Abnormal Control Room radiation levels are adequately bounded by EAL RA3.1. The remaining list of OC dependent rooms in Tables R-3 and H-2 is a list of rooms where actions are absolutely required to be performed to move the plant from normal operations through cool down and to achieve and maintain cold shutdown. These areas are not areas requiring entry to solely meet surveillance requirements. This does not include actions called out for in the procedures that are not absolutely needed to move the plant into and maintain cold shutdown (e.g. temperature adjustment turbine lube oils systems, opening of system drains, RHR ‘hardening’ actions, surveillance/administrative tasks). In order to transition from normal operations (OCs 1 and 2) to hot shutdown (OC 3) the only location required is the Control Room since the plant is able to be placed into hot shutdown from the control room without the need for any other room entry.




The remaining rooms/areas shown in in Tables R-3 and H-2 were determined based on the discussion section above and OCs 1 and 2 were excluded. Therefore, the analysis starting point was a review the steps of HC.OP-IO.ZZ-0004(Q) “SHUTDOWN FROM RATED POWER TO COLD SHUTDOWN" and HC.OP-AB.ZZ-0000(Q) REACTOR SCRAM flow chart. The analysis then branched to other procedure steps described in the IO and AB.ZZ-0000 if those procedures steps were not excluded from consideration as noted in the discussion section (see reviewed references). The analysis concluded that Reactor Building areas and elevations listed in in Table HA5 contain equipment that must be manipulated in OC 3 to achieve and maintain cold shutdown. OCs 4 and 5 are also included for these areas to account for swapping loops for long term operation of shutdown cooling. Table R-3 & H-2 Bases A review of station operating procedures identified the following OC dependent in-plant actions and associated areas that are required for normal plant operation cooldown or shutdown: Procedure step Building Elevation Room #(s) OC IO-4 5.1.15.Q RB 102 4307 3,4,5 Restore power to AB-HV-F020/F021; AB-HV-F001/F002; BG-HV-F031 SOP-BC-0002 5.2.21 RB 77 4201 3,4,5 Electrically arm valve F008 logic to permit valve opening 5.2.44 RB 54 4113 / 4109 3,4,5 Lower Rx water level; only IF normal RWCU blowdown to condenser OR F040/F049 to RW not available 5.2.45 or 5.2.46 RB 102 4328 / 4322 3,4,5 Raise Rx water level




Results:

Table R-3/H-2 Safe Operation & Shutdown Rooms/Areas Bldg. – Ele. Rooms/Areas OPCON

Applicability


RB 77’ 4201 (10B242 MCC) 3, 4, 5



Reviewed references:

1. HC.OP-IO.ZZ-0004 SHUTDOWN FROM RATED POWER TO COLD SHUTDOWN

2. HC.OP-AB.ZZ-0000FC REACTOR SCRAM FLOW CHART 3. HC.OP-SO.AB-0001 MAIN STEAM SYSTEM OPERATION 4. HC.OP-SO.AD-0001 CONDENSATE SYSTEM OPERATION 5. HC.OP-SO.AC-0001 MAIN TURBINE OPERATION 6. HC.OP-SO.AE-0001 CONDENSATE SYSTEM OPERATION 7. HC.OP-SO.BC-0001 RESIDUAL HEAT REMOVAL SYSTEM OPERATION 8. HC.OP-SO.BC-0002 DECAY HEAT REMOVAL OPERATION 9. HC.OP-SO.BC-0003 RHR ALTERNATE COOLING MODES 10. HC.OP-SO.CG-0001 CONDENSER AIR REMOVAL SYSTEM OPERATION 11. HC.OP-SO.BG-0001 REACTOR WATER CLEANUP SYSTEM OPERATION 12. HC.OP-AB.RPV-0003 RECIRCULATION SYSTEM/POWER OSCILLATIONS 13. HC.OP-AB.RPV-0009 SHUTDOWN COOLING 14. HC.OP-AB.HVAC-0002 CONTROL ROOM ENVIRONMENT 15. HC.OP-SO.SA-0001 REDUNDENT REACTIVEITY CONTROL SYSTEM

OPERATION 16. HC.CH-SO.AX-0001 HYDROGEN AIR INJECTION SYSTEM OPERATION