i southern edisondwight california edisonvice president · 2012. 11. 18. · i southern...

I SOUTHERN CALIFORNIA EDISONDwight E. EDISONVice President

An EDISON INTERNATIONAL"' Company

August 2, 2001

U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject: Docket Nos. 50-361, and 50-362 Facility Change Report San Onofre Nuclear Generating Station Units 2 and 3

Gentlemen:

This letter transmits the Facility Change Report required by 10 CFR 50.59(b)(2) for San

Onofre Units 2 and 3 for the period from May 9, 1999 through the end of the Unit 3,

Cycle 11 refueling outage (February 3, 2001). This report (Enclosure 1) provides a

summary of the facility changes, procedure changes, and any tests and experiments,

including a summary of the safety evaluations performed for each change. The scope

of this report is based on an extensive review of plant records, and all 50.59

evaluations that have been identified for the time period above are included in this

report. Complete facility change documentation is available onsite.

Enclosure 2 provides a report on commitment changes made per NEI "Guidelines for

Managing NRC Commitments."

If you would like any additional information, please feel free to contact Mr. J. L.

Rainsberry at (949) 368-7420.

Sincerely,

Enclosures

cc: E. W. Merschoff, Regional Administrator, NRC Region IV

C. C. Osterholtz, NRC Senior Resident Inspector, San Onofre Units 2 & 3

J. E. Donoghue, NRC Project Manager, San Onofre Units 2 and 3

S. Y. Hsu, Department of Health Services, Radiologic Health Branch

P. O. Box 128 San Clemente, CA 92674-0128 949-368-1480 Fax 949-368-1490

ENCLOSURE 1

SAN ONOFRE NUCLEAR GENERATING STATION

UNITS 2 AND 3

FACILITY CHANGE REPORT

FOR THE PERIOD FROM MAY 9, 1999 THROUGH FEBRUARY 3, 2001

SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 FACILITY CHANGE REPORT

CONTENTS

SECTION

1 .....................

2.......................

3 . ...... ...... ..... ...

4. ..... ....... .... ....

5 .....................

6 ..................... ....

7 .........................

8 .......................

9......................

10 . . . . . . . . . . . . . . . . . . . . . . . .

1 1 . . . . . . . . . . . . . . . . . . . . . . . .

12 . . . . . . . . . . . . . . . . . . . . . . . .

DESCRIPTION

Minor Modification Packages (MMPs), Facility Change Evaluations (FCEs), Design Change Packages (DCPs) and Temporary Facility Modification (TFMs)

Abnormal Alignment/Evolutions

Updated Final Safety Analysis Report (UFSAR) Changes

Updated Fire Hazards Analysis (UFHA) Changes

Procedure Changes

Licensee Controlled Specifications (LCS) Changes

Technical Specification (TS) Bases Changes

Non-Conformance Report (NCR) Safety Evaluations

Barrier Program Evaluations

Offsite Dose Calculation Manual (OCDM) Revisions

Control of Core Protection Calculator (CPC) Constants and Core Operating Limit Supervisory System (COLSS) Addressable Constant Changes

Field Change Notices (FCNs)

I

SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 FACILITY CHANGE REPORT

CONTENTS (Continued)

SECTION

13 . . . . . . . . . . . . . . . . . . . . . . . .

14 . . . . . . . . . . . . . . . . . . . . . . . .

15 . . . . . . . . . . . . . . . . . . . . . . . .

16 . . . . . . . . . . . . . . . . . . . . . . . .

DESCRIPTION

Control of Work Storage Areas in Protected Area (Combustible Controls)

Action Request (AR) Safety Evaluations

Design Change Notices (DCNs)

Software Changes (Software Modifications and Installations)

11

SECTION 1


UNITS 2 AND 3

FACILITY CHANGES IMPLEMENTED FOR THE PERIOD FROM MAY 9, 1999 THROUGH FEBRUARY 3, 2001

Facility Change Evaluations (FCEs) Minor Modification Packages (MMPs)

Design Change Packages (DCPs) and Temporary Facility Modifications (TFMs)

1-1


UNITS 2 AND 3

FACILITY CHANGES

1-2

Facility Change Evaluations (FCE) 2-00-002 Rev. 0 and FCE 3-00-003 Rev 0

Title: Unit 2/3 Cycle 11 Core Reload

Description:

This FCE examines the following fuel-related design changes for Unit 2 Cycle 10.

1. Replacement of Nuclear Fuel 2. Minor Mechanical Design Changes

The Cycle 10 reload batch size was 92 Batch N assemblies. The reload batch for Cycle 9 (Batch L) also consisted of 100 fuel assemblies. The N fuel assembly upper end fitting assembly holddown spring design differs from that previously used and the perpendicularity and the conditions of its ends are more tightly controlled. This was done to ease fabrication of upper end fitting assemblies, specifically, insertion of the guide posts through the preloaded springs. This was the only mechanical design change made to the fuel assemblies for Cycle 11.

Safety Evaluation:

A systematic review of each of the appropriate events in Updated Final Safety Analysis Report (UFSAR) Chapters 3, 6, and 15 was contained in the Reload Analysis Report (RAR) which was part of this design modification package. In all cases, the conclusions were that the new core and fuel assembly modifications did not have an adverse or non-conservative impact on any safety analysis.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the UFSAR, did not increase as a result of the Unit 2/3 Cycle 11 core reload modifications. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The Unit 2/3 Cycle 11 core reload modifications had no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications; thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-3

Minor Modification Package (MMP) 2-88-050 Rev. 2

Title: Letdown and Charging System Valves Description:

Letdown and charging system Kerotest valves were replaced with Anchor Darling valves with live load stem packing, to ensure no leakage. These valves are used when maintenance is performed on the systems.

Safety Evaluation:

Since valve S21201 MU991 was installed into the letdown line a thermal fatigue analysis was performed, which confirmed the new valves meet the system design requirements. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. These valve replacements have no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-4

MMP 2/3-6716.OOSJ Rev. 0

Title: Plant Monitoring System (PMS) and Core Operating Limit Supervisory

System (COLSS) Backup Computer (CBC)

Description:

A new computer processing unit (CPU) board was installed to replace the multiboard processors.

Safety Evaluation:

This change allows the PMS to run with a reduced power requirement at a faster speed and more potential computing ability. The new CPU board installation operates identically to the previous system. The Control Room display information was changed by providing the COLSS indicators in the Control Room monitors, which is an improvement for the operators. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This new CPU has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-5

MMP 3-6806.OSM Rev. 0

Title: Main Steam Safety Valve Ring Modifications

Description:

The Main Steam Safety Valve ring settings were changed and the nozzle ring and guide rings were replaced.

Safety Evaluation:

The resultant new ring settings were to ensure that the Safety Valves are capable of their design rated lift at 3% accumulation. These new ring settings and replacement rings increased the blowdown of the Safety Valves. However, the effects of increased blowdown do not impact the design function of the Main Steam Safety Valves (MSSVs). The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. These Main Steam Safety Valve ring settings, nozzle ring, and guide ring changes have no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical

Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-6

MMP 3-6811.OOSM Rev. 0

Title: Component Cooling Water (CCW) System Valve MU138

Description:

A replacement CCW valve MU138 was installed.

Safety Evaluation:

This new valve has a reduced sized internal configuration (trim). The maximum flow rate capacity of CCW through this valve is 1200 gallons per minute, which is equal to that of the original valve's trim with MU1 38 throttled. The replacement valve was evaluated to ensure that it did not alter the loading on the CCW system, whether by volume flow rate requirements, or thermal loading. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. Replacement of this CCW system valve has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-7

MMP 2&3-6858.OOSN Rev. 0

Title: Low Pressure Safety Injection (LPSI) Pumps/Motor Separation

Description:

To accommodate improved cooling to prolong pump seal life, the LPSI pumps were

modified to separate the pump element from the motor by an additional 15". New pump

shaft and spacer couplings were added to couple the pump element to the motor, and a

new bearing was added between the pump upper casing, and lower motor journal bearing. Also, a seal was replaced.

Safety Evaluation:

This modification resulted in lower LPSI pump seal temperatures, prolonging the pump

seal life. There is no negative effect on the pump hydraulic performance. The total

developed head and impeller dynamics are bounded within the original design

conditions. The probability of occurrence or the consequences of an accident, or

malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The

possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This LPSI pump

change has no effect on either the existing Limiting Conditions for Operation or the

Surveillance Requirements in the Technical Specifications: thus, the margin of safety

as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-8

MMP 2&3-6863.01SN Rev. 0

Title: Pressurizer Nozzle Cap

Description:

A 6 inch pipe cap was replaced with a 6 inch weld neck flange and blind flange on a pressurizer spare 6 inch nozzle. The blind flange will be removed when the containment spray pump is used during shut down as a reactor coolant system vent to provide over pressure protection.

Safety Evaluation:

Replacing the welded pipe cap with a blind flange provides a potential leakage path from the pressurizer (weld vs. bolted connection). This pressurizer vent has been analyzed to meet ASME Section III, Class1 requirements and San Onofre has not experienced any leakage through these type of joints. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This nozzle cap replacement has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-9

MMP 2&3-6891.OOSM Rev. 0

Title: Full Flow Condensate Polisher Demineralizer (FFCPD) Effluent Sample Line

Description:

An oxygen analyzer and recorder were added to monitor the dissolved oxygen concentration of the condensate discharged from the FFCPD. This required routing quarter inch tubing from the FFCPD sample header to the new oxygen analyzer.

Safety Evaluation:

The new sample line for this oxygen analyzer and recorder does not interface with or

impact the performance of any component or system important to safety. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This effluent sample line has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in

the bases for the Technical Specifications was not reduced as a result of this change.

1-10

MMP 2&3-6908.OOSJ Rev. 0

Title: Printed Circuit Board for the Process and Effluent Radiation Monitoring

System (PERMS)

Description:

A printed circuit board for the Process and Effluent Radiation Monitoring System (PERMS) was replaced.

Safety Evaluation:

This new printed circuit board was required since the previous printed circuit board was

no longer available and specific components on the printed circuit board which required

repair were also not available with the proper characteristics. This board is used in the PERMS count rate analog circuitry. The new printed circuit board only affects the

PERMS and does not affect the plant operability or function. The probability of

occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report

(UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was

not created as a result of this change. This replacement printed circuit board has no

effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in


1-11

DCP 2&3-6990.OOSN Rev. 0

Title: Reactor Coolant Pump Baffle

Description:

The reactor coolant pump's baffle configuration was modified by adding a tapered ring in the gap between the rotating baffle and the pump shaft, to resolve possible baffle failure mechanisms.

Safety Evaluation:

The taper ring was tested and analyzed to simulate the interaction between it and the reactor coolant pump shaft and baffle. The baffle cap screw, including the locking cup, function, and design basis are unaffected by the changes. The modification did not affect the pump hydraulics or the mechanical operation parameters, and seismic qualification remains unchanged. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. Adding this tapered ring in the gap between the reactor coolant pump rotating baffle and the pump shaft has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-12


UNITS 2 AND 3

TEMPORARY FACILITY MODIFICATIONS

1-13

Temporary Facility Modification (TFM) Number TFM-3-01-BBA-001 Rev. 0

Title: Temporary Video Camera in Containment

Description:

A video camera was temporarily installed near the lower lube oil reservoir sight glass of reactor coolant pump motor 331201 MMOO1. The camera was used to monitor the lube oil reservoir sight glass for possible step changes in oil level during operation.

Safety Evaluation:

Containment cleanliness requirements, seismic concerns, and fire loading changes resulting from this temporary change were evaluated and determined to be acceptable. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this temporary change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This temporary change had no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-14

Temporary Facility Modification Numbers and 2-00-SBB-1. TFM-2-01 -SBB-1, and TFM-2-01-SSB-2

Title: Simulated Signal Input for Control Element Assembly (CEA) 40 Reed

Switch Position Transmitter (RSPT)

Description:

To eliminate intermittent nuisance sensor failure alarms for CEA 40, caused by a faulty field signal, a zener diode resistor combination was installed in cabinet 2L091 to simulate a full out position indication for CEA 40, and a 10 VDC source was connected to simulate CEA position for CEA 40.

Safety Evaluation:

These modifications affected only the CEA position indication, not the method by which the CEAs are held or moved. Failure of a CEA position indication is not assumed in the UFSAR as the cause of any accident. CEA 40 is in the shutdown bank of control rods, and is required to be fully withdrawn from the core during operation. CEA 40 is not one of the target CEAs that provide an input to the Core Protection Calculators (CPCs). These changes did not impact the CPCs or the Control Element Assembly Calculator's ability to automatically generate any reactor trip signals. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. These temporary changes had no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-15

Temporary Facility Modification Number TFM-2-99-BBA-1

Title: Unit 2 Control Element Assembly Calculator (CEAC) Sensor Failure Annunciator Alarm

Description:

The Unit 2 CEAC Sensor Failure Annunciator Alarm (2UA0056C52) in the Control Room was disabled to prevent nuisance alarms caused by a defective Reed Switch Position Transmitter (RSPT) signal to CEA 40.

Safety Evaluation:

Indication of CEAC sensor failure is available to the Control Room operators on the CPC/CEAC remote operator module. The CPC/CEAC was still capable of automatically initiating a reactor trip signal to the reactor protection system. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this temporary change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This temporary change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-16

Temporary Facility Modification Number TFM-2-00-BBA-1 Rev. 0

Title: Reactor Coolant Pump (RCP) 2P003 Intergasket Leakage Alarm

Description:

The RCP 2P003 Intergasket Leakage Alarm was extinguished by shorting the input from pressure switch 2PSH-0231.

Safety Evaluation:

The alarm was caused by leakage past the inner casing of the pump, causing pressure buildup in the annulus between the inner and outer casings when the leakoff drain line to the Reactor Coolant Drain Tank (RCDT) was isolated. Defeating the 2P003 intergasket leakage alarm allows the intergasket leakage alarm for the remaining reactor coolant pumps to remain operable until the 2P003 gasket can be replaced at the next refueling outage. Monitoring of the RCP 2P003 outer gasket will be performed by the water inventory balance Surveillance which is performed every 72 hours. This monitoring is acceptable since the gasket joint is a static seal, and any changes in gasket integrity occur slowly over an extended time when the Reactor Coolant System (RCS) is at steady state conditions. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this temporary change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This temporary change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

1-17

Temporary Facility Modification Number TFM-2-00-FLA-1

Title: Main Feedwater Pump Turbine (MFWPT) Thrust Bearing High Wear Trip

Circuits

Description:

To reduce the possibility of a spurious MFWPT trip being caused by erratic indication

of high thrust bearing wear, the Thrust Bearing High Wear Monitor (2NIT-4533F) trip

circuit was removed from service. In addition, Channel "B" of 2NIT-4533F was turned off.

Safety Evaluation:

These changes left the turbine with reduced tripping protection because both Channel "A and B" trip circuits were being defeated. However, Channel "A" alarm circuit was

available to still function as normal and provide the Control Room with pre-trip alarms

should the axial movement or actual thrust bearing wear reach 20 mils. The plant operators would then take action required to evaluate and correct the cause of the

problem or, if necessary, trip the MFWPT to prevent further damage to it or other plant

equipment. If the turbine had excessive axial shaft movement or high vibration, the Vibration Monitoring and Trip Circuit would detect this movement and provide alarms

or, if severe enough, trip the turbine. The feedwater pump, turbine and turbine control

system are Quality Class Ill, Seismic Category II, are not Important-To-Safety, and perform no function in the safe shutdown of the plant. Defeating the turbine thrust trip

reduced the likelihood of a spurious trip caused by erroneous high thrust wear signals

from the remaining good probe. The high thrust trip of the MFWPT is not part of the

design bases for the feedwater system as defined in UFSAR Section 10.4.7.1. The

defeated high thrust trip was for equipment protection only and does not perform a plant protection function. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in

the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this

temporary change. The possibility of either an accident or malfunction of a different

type than previously evaluated in the UFSAR was not created as a result of this

change. This temporary change had no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus,

the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5059dcpetc#1

1-18

SECTION 2


UNITS 2 AND 3

FACILITY CHANGES IMPLEMENTED FOR THE PERIOD FROM MAY 9,1999 THROUGH FEBRUARY 3, 2001

Abnormal Alignment/Evolutions

2-1

Abnormal Alignment 2-99-76:

Title: 2HV6497, Salt Water Cooling (SWC) From Component Cooling Water (CCW) Heat Exchanger HX E001 To Outfall

Description:

Valve 2HV6497 was prevented from inadvertently closing automatically upon transfer of the 4 kV bus to the opposite Unit due to an identified circuit deficiency. This was accomplished by opening breaker 2BY35 to the 2HV6497 power supply. This alignment was in effect from 6/25/99 to 8/13/99.

Safety Evaluation:

This alignment enhanced plant safety by preventing inadvertent closure of valve 2HV6497. The safety function of the valve is to open or remain open on a Safety Injection Actuation Signal (SIAS). The valve is designed to close automatically upon SWC pump breaker opening. This automatic close feature was disabled by the alignment. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change because the valve was failed open to its safety related position. There is no requirement in the Technical Specification Bases for the valve to be closed during operation of the SWC system. The inservice testing program does not test the closing function of the valve, as it is not a required safety function.

2-2



Description:

Valve 2HV6495 was prevented from inadvertently closing automatically upon transfer of the 4 kV bus to the opposite Unit due to an identified circuit deficiency. This was accomplished by opening breaker 2BZ31 to the 2HV6495 power supply. This alignment was in effect from 6/25/99 to 8/28/99.

Safety Evaluation:


2-3


Title: 2HV6497, Salt Water Cooling (SWC) From Component Cooling Water (CCW) Heat Exchanger HX EO01 To Outfall

Description:

Valve 2HV6497 was prevented from inadvertently closing automatically upon transfer of the 4 kV bus to the opposite Unit due to an identified circuit deficiency. This was accomplished by opening breaker 2BY35 to the 2HV6497 power supply. This alignment was in effect from 8/13/99 to 9/10/99.

Safety Evaluation:


2-4


Title: 2HV6495, SWC From CCW HX E002 To Outfall

Description:

Valve 2HV6495 was prevented from inadvertently closing automatically upon transfer of the 4 kV bus to the opposite Unit due to an identified circuit deficiency. This was accomplished by opening breaker 2BZ31 to the 2HV6495 power supply. This alignment was in effect from 8/28/99 to 9/2/99.

Safety Evaluation:


2-5



Description:

Valve 3HV6495 was prevented from inadvertently closing automatically upon transfer of the 4 kV bus to the opposite Unit due to an identified circuit deficiency. This was accomplished by opening breaker 3BZ31 to the 3HV6495 power supply. This alignment was in effect on 6/25/99 from 0302 to 0355.

Safety Evaluation:


2-6



Description:

Valve 3HV6497 was prevented from inadvertently closing automatically upon transfer of the 4 kV bus to the opposite Unit due to an identified circuit deficiency. This was accomplished by opening breaker 3BY35 to the 3HV6497 power supply. This alignment was in effect from 6/25/99 (0302) to 6/25/99 (0355).

Safety Evaluation:


2-7


Title: Gather Data To Support Operability of Radiation Monitor 2RT7870

Description:

Radiation Monitor 2RT7870 was performance tested during elevated condensate temperatures experienced during heat treats of the circulating water system performed at 100% power. This alignment was in effect on 6/30/99 from 1420 to 1430.

Safety Evaluation:

The radiation monitor itself remained operable during the test. Heat tracing on the sensing lines was declared inoperable but remained functional. The test raised the heat tracing sepoint to a maximum of 170 degrees F to simulate higher temperature of

the sample flow. The test lowered the automatic pump setpoint that initiates operation of the high/mid range sample pump in order to obtain sample flow data with the high/mid range pump running concurrently with the low range pump. The probability of

occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this test. The test was designed so that the 130 degrees F design limit for the air dryers and radiation monitor detectors was not exceeded. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this test. This test

had no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change. Radiation monitors are listed in the Offsite Dose Calculation Manual (ODCM) and the Licensee Controlled Specifications (LCS).

2-8



Description:

Radiation monitor 3RT7870 was performance tested during elevated condensate temperatures experienced during heat treats of the circulating water system performed at 100% power. This alignment was in effect on 6/24/99 from 0846 to 1640.

Safety Evaluation:

The radiation monitor itself remained operable during the test. Heat tracing on the sensing lines was declared inoperable but remained functional. The test raised the heat tracing sepoint to a maximum of 170 degrees F to simulate higher temperature of the sample flow. The test lowered the automatic pump setpoint that initiates operation of the high/mid range sample pump in order to obtain sample flow data with the high/mid range pump running concurrently with the low range pump. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this test. The test was designed so that the 130 degrees F design limit for the air dryers and radiation monitor detectors was not exceeded. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this test. This test had no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change. Radiation monitors are listed in the Offsite Dose Calculation Manual (ODCM) and the LCS.

2-9



Description:

Radiation monitor 3RT7870 was performance tested during elevated condensate temperatures experienced during heat treats of the circulating water system performed at 100% power. This alignment was in effect from 7/28/99 to 7/29/99.

Safety Evaluation:

The radiation monitor itself remained operable during the test. Heat tracing on the sensing lines was declared inoperable but remained functional. The test raised the heat tracing sepoint to a maximum of 170 degrees F to simulate higher temperature of the sample flow. The test lowered the automatic pump setpoint that initiates operation of the high/mid range sample pump in order to obtain sample flow data with the high/mid range pump running concurrently with the low range pump. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this test. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this test. This test had no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change. Radiation monitors are listed in the Offsite Dose Calculation Manual (ODCM) and the LCS.

5059abn

2-10

SECTION 3


UNITS 2 AND 3

FACILITY CHANGES IMPLEMENTED FOR THE PERIOD FROM MAY 9,1999 THROUGH FEBRUARY 3,2001

Updated Final Safety Analysis Report (UFSAR) Changes

3-1

UFSAR 2/3 Change REQUEST No. 18, Section 05.04

Title: Reactor Coolant System (RCS) Flow Restricting Orifices

Description:

UFSAR Section 5.04 was revised to 1) provide additional design information on the RCS flow restricting orifices and 2) address the standpipe type of flow restricting orifices for the lower pressurizer level instrument nozzles, to be consistent with information provided in Design Bases Document DBD-SO23-360.

Safety Evaluation:

This UFSAR change adds information on RCS flow restricting orifices into the UFSAR which is already provided in DBD-SO23-360 and involves no physical change. Technical Specification Bases Sections 3.4.13 and 3.4.15 are not affected by this

UFSAR change. The probability of occurrence or the consequences of an accident, or

malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The

possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This UFSAR change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety

as defined in the bases for the Technical Specifications was not reduced as a result of this change.

3-2

UFSAR 2/3 Change REQUEST No. 31, Section 09.02 (AR 000301907)

Title: Condensate Storage System

Description:

The UFSAR change to Section 9.2.6 reflects the current as-built design of systems connected to the Condensate Storage System. Specifically, the change involves the

deletion of the Nuclear Service Water Tank T-1 04 and inclusion of the Turbine Cooling Water Tank T-050 in Section 9.2.6.1 .1.

Safety Evaluation:

The as-built connected systems to the Condensate Storage System have been evaluated in the calculation M-0050-017, and adequate condensate volume margin has been allocated to account for associated system leakages and outflows.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety

Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This UFSAR change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

3-3

UFSAR 2/3 Change REQUEST No. 15. Section 03.05 (AR 000501065)

Title: Tornado Missile Protection

Description:

UFSAR Table 3.5-13 was revised to update the tornado missile damage probabilities. The total annual probability of damage to exposed critical components increased from 0.2E-7 per unit to 0.4E-7 per unit. The approved acceptance limit for the total annual probability of damage to exposed critical components, 1 E-7 per unit, was unchanged. The probability of tornado missile damage to exposed critical components is a function of the site inventory of potential tornado missiles. Reported damage probabilities in UFSAR Table 3.5-13 were previously based on a potential missile inventory that was established in 1990, when all three SONGS units were operating. Unit 1 decommissioning work increases the site inventory of potential missiles. This increased inventory produces an increase in the probability of damage to exposed critical components in Units 2 and 3.

Safety Evaluation:

NRC approval of the acceptance limit for probability of damage to exposed critical components, 1.01E-7 per year per unit, was obtained with approval of License Amendments 148 (Unit 2) and 140 (Unit 3). The revised total probability of damage to exposed critical components, 0.4E-7 per year per unit, is less than the NRC-approved acceptance limit.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This UFSAR change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

3-4

UFSAR 2/3 Change REQUEST No. 38. Section 09.03 (AR 000601165) UFHA 2/3 Change REQUEST No. 15 UFHA 2/3 Change REQUEST No. 16

Title: Upgraded Sump Level Controller and Combustible Fire Loading Changes

Description:

UFSAR Section 9.3.3.2.3 was revised to describe the design and operation of new non-radioactive sump level controls and alarm switches. A new ultrasonic style level control and alarm switch have been installed in some locations to improve operational reliability. This change required the installation of additional cable from the associated sump to the motor control center (MCC) for each sump. The additional cable affects the Updated Fire Hazard Analysis (UFHA) for Unit 2 by increasing the combustible fire loading for that area.

Safety Evaluation:

The increase in combustible fire loaded caused by the additional cables is within design limits. The design bases of the non-radioactive sump level control system is unchanged.


3-5


Title: Erbia Fuel Thermal Conductivity and Control Element Assembly (CEA) Ejection Event Analysis

Description:

UFSAR Section 15.10.4.3.2 was revised to reflect the most recent Cycle 10 analyses for the CEA ejection event for both Unit 2 and Unit 3. The re-analysis was done to correct minor errors in the erbia fuel thermal conductivity and in the Departure from Nucleate Boiling Ratio (DNBR) calculation.

Safety Evaluation:

The Cycle 10 transient analyses continue to meet the fuel energy deposition acceptance criteria. The radiological dose analysis is unaffected, thus the dose consequences are unchanged.


3-6

UFSAR 2/3 Change REQUEST No. 21! Section 09.01 (AR 001001374)

Title: Spent Fuel Pool (SFP) Cooling Operation During Full Core Off load

Description:

This change corrected statements made in UFSAR Section 9.1 regarding SFP cooling.

For the maximum refueling full core offload heat load case, two SFP cooling pumps and

heat exchangers (instead of one) will be needed to maintain the SFP temperature at or below 160 deg F.

Safety Evaluation:

The maximum refueling full core offload heat load case is not a design base scenario and the heat load (43 MBTU/Hr) is enveloped by the maximum abnormal heat load (worst case scenario; 51.3 MBTU/Hr). The change does not impact the safety function

of the system and it does not change any accident evaluated previously in the Safety Evaluation Report.

The probability of occurrence or the consequences of an accident, or malfunction of

any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of

either an accident or malfunction of a different type than previously evaluated in the

UFSAR was not created as a result of this change. This UFSAR change has no effect

on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in


3-7


Title: Containment Passive Heat Sink Inventory

Description:

UFSAR Section 6.2 was updated to reflect the change in containment passive heat sink inventory. An increase in the amount of galvanized steel inside containment increased the containment passive heat sink inventory.

Safety Evaluation:

The current Large Break LOCA ECCS analysis of record bounds the current conditions. The impact of the new heat sink inventories on the large break LOCA ECCS verification analysis was offset by an allowed value of initial containment pressure. The containment pressure used in the analysis is equal to the Limiting Condition of Operation (LCO) value minus the existing total loop uncertainty (TLU) of -0.9 PSIG (13.3 PSIA). This value is fully consistent with current design values, and there is no actual change to the plant design or to the way in which the plant is operated.


3-8


Title: Chemical and Volume Control System (CVCS) Letdown Line Isolation

Description:

UFSAR section 9.3 was revised to provide a description for isolating the CVCS letdown line based on a high temperature downstream of the regenerative heat exchanger. The letdown isolation function of the temperature controlled isolation valves 2(3)TV0221 and 2(3)TV9267 had been credited in analysis of letdown line breaks outside of containment.

Safety Evaluation:

Crediting the isolation of the CVCS letdown line in response to a high temperature sensed downstream of the regenerative heat exchanger provides for faster isolation of the break. Letdown line isolation is completed at fifty seconds after a postulated guillotine pipe break outside containment, which is faster than the 1800 seconds previously evaluated and summarized in table 15.6-1 of the UFSAR. Crediting this isolation will not result in any system being operated outside of its design or testing limits nor will it cause a change to any system interface.


3-9


Title: Purification and Deborating Ion Exchangers

Description:

UFSAR section 9.3.4 was modified to allow any of the purification ion exchangers to be loaded and operated as a deborating ion exchanger. This change allows for additional flexibility in operating the ion exchangers in either mode during the fuel cycle.

Safety Evaluation:

The ion exchangers are of similar design and construction and can be utilized interchangeably for purification or deboration based on plant and fuel cycle requirements. The pressure boundary, the flow characteristics, the design temperature and pressure are similar. No additional loads are imposed, no protection features are modified or deleted, no system redundancy or independence is reduced, no frequency of operation is increased, and no testing requirements are changed.


3-10


Title: Temperature Element Location in Emergency Cooling Units

Description:

UFSAR Section 9.4.2.2.3 was revised to reflect the as-built configuration of the high temperature elements in the Emergency Core Cooling System (ECCS) and Component Cooling Water (CCW) pump room emergency cooling units (ECUs). Some of the ECUs were found to have temperature elements (TEs) located between the cooling coil and the fan inlet instead of the fan discharge as previously indicated in the UFSAR.

Safety Evaluation:

Locating the temperature elements between the cooling coil and fan inlet of the ECU will not expose the ECCS and CCW equipment in the pump rooms to excessive temperature beyond their environmental qualification, and their safe shutdown and mitigation functions will not be adversely affected should an ECU fail unnoticed. Furthermore, the emergency cooling system was designed such that a single failure of an active component will not result in a complete loss of ECCS or CCW pump function (UFSAR 9.4.3.2.3). Therefore, the proposed activity is bounded by the existing design and safety analyses.


3-11


Title: Unit 2 Trains A and B Emergency Room Cooler Capacity Verification

Description:

This UFSAR change updates Section 9.4.3.2.2.2, Tables 9.4-7, 8 and 10 by adding the minimum required cooling coil capacity values for the Unit 2 Train A/B Emergency Room Coolers as documented in calculation M-075-052.

Safety Evaluation:

Calculation M-0075-052, Unit 2 Loops A and B Emergency Room Cooler Capacity Verification, documents adequate cooling of their respective area/rooms is available with the minimum chilled water flow rates of these cooling units based on the heat load requirements as specified in the UFSAR. This analysis also accounts for reductions in cooling coil capacities due to the differences between pre-operational test data and nominal design flow rates.


3-12


Title: Recalculation of Containment Temperature and Pressure Response During Design Bases LOCA Events

Description:

UFSAR Section 3.11 was revised to document recalculation of the containment transient pressure and temperature response to a spectrum of design basis loss of coolant accidents (LOCAs). The re-analysis was originally undertaken as a part of the reduction in Tcold project for SONGS Units 2 and 3 and includes updated plant design and performance input data, consideration of instrumentation total loop uncertainties (TLUs), and updated passive heat sink data. In addition, the revised analysis includes new mass and energy release data using revised large break LOCA methodology reviewed and approved by the NRC in 1985. Mass-energy release calculations confirmed that the original Tcold, still permitted by Technical Specification LCO 3.4.1, was more limiting for containment pressure-temperature response; therefore, the new analysis continues to be based on a maximum Tcold of 560 degees F.

Safety Evaluation:

The revision of the containment post-LOCA pressure-temperature response calculation has resulted in a change in the worst case LOCA event from the standpoint of short-term peak pressure from the previously identified double-ended suction leg slot break LOCA with diesel generator failure to the double-ended discharge leg slot LOCA with diesel generator failure. The peak containment post-LOCA pressure is reduced from 55.1 psig to 45.9 psig. The worst case LOCA event from the standpoint of short-term peak vapor temperature is now the double-ended hot leg slot break LOCA with diesel generator failure, and the peak vapor temperature is reduced from 295F to 268F. The limiting LOCA for long-term containment cool-down is now the double-ended hot leg slot break LOCA with diesel generator failure, rather than the double-ended suction leg slot break LOCA previously identified. The lower peak containment pressure and temperature reduce the consequences of the design basis LOCA event, and increase the margin of safety. The radiological consequences of design basis LOCA events in containment remain unchanged from that based on the previous analysis of record. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This UFSAR change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

3-13


Title: CVCS Malfunction (Boron Dilution) Event Description

Description:

The description of the CVCS Malfunction (Boron Dilution) event in the UFSAR Section 15.10.4.1.4 and DBD Section 4.1.13 was changed to be consistent with the safety analysis. The change reduces the RCS volumes credited in the safety analysis from 430,000 Ibm in Mode 5 to 387,000 Ibm.

Safety Evaluation:

This change simply sets a more restrictive initial condition for the safety analysis as described in the UFSAR and the DBD. No change in actual RCS volume was proposed. There was no physical change in design, material, or in construction of the plant, nor any change in the way the plant is operated. An unplanned boron dilution would be terminated before shutdown margin would be lost, therefore, no fission product barriers would be affected.


3-14

UFSAR 2/3 Change REQUEST No. 29. Section 6.2

Title: Containment Systems

Description:

UFSAR Section 6.2.1.1.3.6 was revised to include appropriate information from

calculation N-4080-009 Revision 01, the calculation of record for the maximum external

differential pressure of the containment structure.

In Q&R 022.34, the NRC stipulated the methodology to produce "an acceptable calculation." This calculation became N-4080-009. Revision 01 includes the minimum

technical specification initial pressure with total loop uncertainties.

Safety Evaluation:

This calculation used a combination of the perfect gas and Dalton's laws, assuming the

containment atmosphere is cooled to the minimum containment spray temperature.

This methodology provides an NRC acceptable negative pressure calculation of the

consequences of inadvertent containment spray operation. The pressure retaining

components remain functional because, based on the calculated external pressure

value of 4.1 pounds per square inch gauge (psig), as shown on Updated Final Safety

Analysis Report (UFSAR), Table 6.2-2, there was no reduction in the safety margin.

The containment design external pressure of 5.0 psig was not exceeded.


any equipment important to safety, previously evaluated in the UFSAR, did not

increase as a result of this change. The possibility of either an accident or malfunction

of a different type than previously evaluated in the UFSAR was not created as a result

of this change. This UFSAR change has no effect on either the existing Limiting

Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical

Specifications was not reduced as a result of this change.

3-15

UFSAR 213 Change REQUEST No. 2, Section 15.10

Title: Pre-Trip Steam Line Break (SLB) Transient/Dose

Description:

U FSAR Section 15.10.1.3.1A Steam Line Break was revised to reflect the Principal Assumptions and inputs and the Results of the Pre-Trip Steam Line Break Transient Analysis NFM-2/3-TA-0002 and Dose Analysis NFM-2/3-DS-0901.

The Pre-Trip Steam Line Break Transient Analysis FM-2/3-TA-0002 and Dose Analysis NFM-2/3-DS-0901 inputs and results have changed to reflect the change in loss of offsite power timing. The reanalysis was in response to a transient analysis methodology change.

Safety Evaluation:

The Pre-Trip SLB is the most limiting non-Loss of Coolant Accident (LOCA) event for the limiting faults category of events. However, the changes in this event do not affect the margin of safety.


3-16


Title: Containment Airflow Value

Description:

UFSAR Section 9.1 Table 9-4.1 Lower Level Circulation Unit Flow Rate was revised to be consistent with the as-built containment airflow value.

Safety Evaluation:

Revising the Lower Level Circulation Unit flow rate from 28,000 to 28,900 cubic ft per minute reflects the original design and as-built value. This change does not affect the Lower Level Circulation Units function, nor does it affect their ability to assist in minimizing hot spots in the containment lower levels.


3-17


Title: Transient Limit Program

Description:

UFSAR Table 3.9-1 was revised to remove insignificant transients, transients no longer valid, and transients redundant to the component cyclic or transient limit program. The following transients were removed from the UFSAR Table 3.9-1 and relocated to the text of paragraph 3.9.1 .1.

a. Power changes b. Normal cyclic variations c. Primary system hydrostatic test d. Primary system leak test

Safety Evaluation:

For power changes and normal cyclic variations, alternate stresses calculated are less than endurance limits. These changes effectively implement the Technical Specification Component Cyclic Transient Program. The capability of the plant to achieve safe shutdown is not affected.


any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This UFSAR change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

3-18


Title: Radiation Protection Design Features

Description:

UFSAR Section 12.3, and Figures 12.3-26, 12.3-27, 12.3-42, 12.3-43, 12.3-54 and

12.3-55 were revised to show the changes to reflect the post accident zoning impact of

safety equipment building doorways to the turbine building and post accident zoning

impact of containment purge inlet penetrations.

These changes are a result of the evaluation of shielding effectiveness that were questioned as part of the UFSAR ReviewNerification Program.

Safety Evaluation:

The radiation zone maps are used to plan operator and maintenance access following an accident. They are a convenient spatial translation of projected dose rate

information given in calculations. There are no errors noted in the calculations. These

drawing changes are required to correct minor differences in dose rates and source

terms, following an accident.





on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

3-19


Title: Radioactive Waste Management

Description:

UFSAR Section 11.1.8.4 was revised to reflect the assumptions, methodology and results of Calculation N-0320-007 Revision 1 as they relate to determining the maximum allowable South Yard Facility contamination limits.

Safety Evaluation:

The quantity of material being decontaminated, as discussed in this Updated Final Safety Analysis Report (UFSAR) section, was chosen to conservatively envelope the maximum amount of material that might be present. Using the results of Calculation N0320-007 Revision 1 as they relate to determining the maximum allowable South Yard Facility contamination limits ensures that all effluent and dose criteria are met.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the UFSAR, did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This UFSAR change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

3-20

UFSAR 2/3 Change REQUEST NO. 3. Section 5.3

Title: Reactor Vessel Surveillance Capsule Details

Description:

UFSAR Tables 5.3-4 and 5.3-5 were revised to show updated lead factors for surveillance capsule locations. Also, Tables 5.3-4, 5.3-5, 5.3.11, and 5.3-12 were

revised to show the capsule removal sequence based on the updated lead factors. Table 5.3-12 was revised to show withdrawal time in Effective Full Power Years (EFPY)

instead of calendar years; a typographical error was corrected in section 5.3.1.6.3.4 concerning a reference section in 10 CFR50 Appendix H, information about the low

leakage fuel design was added, an error was corrected in Table 5.3-4 which listed the

withdrawal schedule of 28 EFPY for the third capsule to be removed. The correct schedule is 24 EFPY.

Safety Evaluation:

This change minimizes the calculational uncertainties in the material condition of the

vessel materials and ensures that the capsule with the proper lead factor is chosen for examination.


any equipment important to safety, previously evaluated in the Updated Final Safety

Analysis Report (UFSAR), did not increase as a result of this change. The possibility of



on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

3-21


Title: RCS Pipe Base Metal

Description:

UFSAR Section 5.4.3 was revised to delete the supplemental inspection requirement and address the long-term corrosion effects on the carbon steel Reactor Coolant System (RCS) pipe base metal.

Safety Evaluation:

This change address the long-term corrosion effects on the carbon steel Reactor Coolant System (RCS) pipe base metal, resulting from the nozzle repair methods, are within RCS design bases for the 40 year plant design life.


3-22

UFSAR 2/3 Change Request No. 3. Section 15.10B

Title: Fuel Handling Accident (FHA) Source Term

Description:

UFSAR Section 15.101B.2.2.2 was revised to present the average fuel rod activity profile used in the FHA dose assessments.

Safety Evaluation:

All accident dose consequences continue to meet acceptable dose criteria. There was no instance where the margin of safety was reduced.


5059ufsar

3-23

SECTION 4


UNITS 2 AND 3


Updated Fire Hazards Analysis (UFHA) Changes

4-1

Updated Fire Hazards Analysis (UFHA) Change Nos. FHA- 7.0-2TB-112, 7.0-2SE-10, 7.0-3TB-7. and 7.0-3SE-1 1 (AR 000201370)

Title: Power Receptacles and PVC Coated Cable

Description:

Power Receptacles were added on the safety equipment building roof to provide a permanent 120 VAC power source and approximately 200 ft of PVC coated ALS cable per unit was installed from the lighting panel to the receptacles.

Safety Evaluation:

The power receptacles were to eliminate the use of extension cords and non-weather plugs to run temporary power from the 480 VAC welding receptacle and 480/120 VAC transformer. Adding a permanent source of 120 VAC power had no affect on the design or function of any safety related equipment or system. However, the addition of

PVC coated cable in the turbine building and safety equipment building roof fire zones

impacted the UFHA. This cable is qualified to the vertical tray flame test requirements of IEEE-383. Addition of the PVC coated cable and resulting combustibles in the turbine building and safety equipment building roof did not cause the maximum allowable to be exceeded as documented in fire area/zones. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was

not created as a result of this change. This additional PVC cable had no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

4-2

UFHA Change Nos. UFHA 2/3-4.0 REV 14 (UFHA change # 7)

(AR 000501066)

Title: Information in the Updated Fire Hazards Analysis (UFHA)

Description:

Estimated fire temperature calculated values and associated methodology were removed from the UFHA.

Safety Evaluation:

Removal of the calculated fire temperature values from UFHA was inconsequential with

respect to the contents of the UFHA and fire protection program features at San Onofre. These values have been reviewed and found to be inaccurate, over-estimated, and overly conservative. The values are not used for design and/or analysis of fire

protection features. Therefore, removal of these calculated values did not delete any critical information from UFHA and/or adversely impact the established fire protection program. This UFHA change did not remove, add, or modify any structures, systems, or component important to safety. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a

result of this change. The possibility of either an accident or malfunction of a different

type than previously evaluated in the UFSAR was not created as a result of this change. Removal of this information from the UFSAR had no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

4-3

UFHA Change Nos. UFHA Change #15. 16 (AR 000601165)

Title: Sump Level Control and Alarm Switches and Cable

Description:

Ultrasonic style level control and alarm switches were installed in the turbine building and other non-radioactive sump locations, and additional cable was installed from the associated sump to the motor control center (MCC) for each sump.

Safety Evaluation:

The level switches are configured such that if an instrument or power failure should occur, the associated control room HI/HI sump level alarm will annunciate in the control room. This design reduced pump wear and tear by not allowing the pumps to run dry. In the case of the turbine sump monitors, the modification improved the performance of the radiation monitor by ensuring accurate sump pump start and stopping. This design change added new cables to the Unit 2 and 3 Turbine Buildings. The combustible loading due to the cable insulation has been evaluated, and the increase in fixed combustible loading remains well below the maximum permissible for this fire area/zone. This change does not create any new fire hazards as previously evaluated in the UFHA. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This additional PVC cable had no effect on either the existing Limiting Conditions for Operation or the

Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

4-4

UFHA 2/3 Change REQUEST No. 66, Section 7.0-2/3AC (AR 000600911)

UFHA 2/3 Change REQUEST No. 11. Section 7.0-2SE

Title: Combustible Fire Loading Change

Description:

UFHA Section 7.0 was revised to provide for the addition of rubber mats in fire area/zones 2-AC-9-8/13/18, 2-AC-50-48/49/50/51/61, 3-AC-50-52/53/54/55 and 2-SE-30-143.

Safety Evaluation:

The increase in combustible loading due to addition of the rubber mats in these areas has not invalidated the bases and results of previously performed fire hazards analyses. The safety related/safe shutdown components/cables located in the area/zones are not adversely impacted due to increase of combustible loading due to addition of rubber mats.



Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFHA was not created as a result of this change. This UFHA change has no effect on

either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5050ufha

4-5

SECTION 5


UNITS 2 AND 3

FACILITY CHANGES IMPLEMENTED FOR THE PERIOD FROM MAY 9,1999 THROUGH FEBRUARY 3,2001

Procedure Changes

5-1

S0123-1-2.5. Revision 6. TCN 6-1:

Title: Design Basis Load Profile for Class 1 E Batteries 2(3)B009 and 2(3)BO1 0

Description:

A revised Design Basis Load Profile for Class 1 E Batteries 2(3)B009 and 2(3)B01 0 was incorporated into the battery service test procedure S0123-1-2.5. The Load Profile was

reduced from 8 hours to 90 minutes because the Shutdown Cooling Valve Inverters 2(3)Y006 and 2(3)Y007 were removed from service.

Safety Evaluation:

The battery service test is performed on an inoperable battery during a refueling outage. The battery is disconnected from operable plant equipment during the testing, so there is no possibility of any plant impact from the maintenance activity. Battery service testing being performed on an inoperable battery isolated from the 125 VDC bus did not impact any required system. Technical Specification SR 3.8.4.7 Bases require that the battery service test length should be as per the duty cycle requirements for the LOVS/SIAS accident profile. The duration of this profile is consistent with the LOVS/SIAS 90 minute profile for the train A and B batteries. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report

(UFSAR), was not increased as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-2

S023-13-13. Revision 7:

Title: Control Element Assembly (CEA) Misalignment Procedure

Description:

Improvements were made to the CEA Misalignment procedure, S023-13-13. The improvements included 1) explicitly allowing CEA insertion to control Axial Shape Index (ASI) and assist with the negative reactivity insertion associated with the downpower which must be performed as a result of the CEA misalignment (e.g., dropped CEA) [note that the procedure already allowed this action - the 50.59 provided further justification based on assumed initial conditions in the analysis and historically expected CPC margin], 2) ensuring the downpower is continued beyond 2 hours if the misalignment is not corrected in that time frame, and 3) allowing one attempt to realign the CEA without assistance if indications support the attempt.

Safety Evaluation:

The misaligned CEA is an anticipated operational occurrence. Requiring a further power reduction of 10% per hour beyond the reduction required by LCS 3.1.105 in the first two hours following a CEA misalignment is a conservative action that will compensate for increasing radial power distortion factors. The procedural changes to expedite retrieving a dropped CEA, including increasing the time frame from 1 hour to 2 hours during which operators are not limited to 3"/minute CEA withdrawal rate, only serve to minimize the effect of xenon redistribution on radial power peaking. If the radial power distortion is minimized, the consequence of a misaligned CEA is reduced. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), was not increased as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-3

S023-3-3.43.36, Revision 3;S023-3-3.43.37, Revision 3;SO23-3-3.43.38. Revision 4:

Title: Engineered Safety Features (ESF) Subgroup Relay Semi-Annual Test

Description:

Subgroup relays K304A, K304B, K306A, and K306B were tested during power operations. The UFSAR (notes 9 and 10 of table 7.3-19) had administratively precluded testing subgroup relays K304A, K304B, K306A, and K306B during power operations in the past in order to prevent a potential Component Cooling Water (CCW) flow perturbation to the Reactor Coolant Pump (RCP) Seals.

Safety Evaluation:

With the proper administrative procedural controls in place, ESF subgroup relay testing is allowed during reactor operation. The required administrative controls include a CCW system alignment such that the train being tested is not aligned to the non critical loop supplying flow to the RCP seals. The proposed testing produces no change in flow to the RCP seals since the train not being tested is supplying flow to the RCP seals resulting in no flow change to the RCP seals. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), was not increased as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-4

S023-XXVII-29.87, Revision 0:

Title: Remote Rolled Plugging of Steam Generator (SG) Tubes

Description:

Stainless steel "fingers" were replaced with low-density polyethylene "fingers" as an installation aid during the removal of a SG tube from service.

Safety Evaluation:

A "rolling" process is used to make the seal between the plug and the SG tube or sleeve wall when installing a tube or sleeve plug. Low-density polyethylene is an

acceptable alternate material to stainless steel for use as an installation aid in the

steam generator roll plugging process. No functional changes were made, and no aspects of the plug qualification were affected.

The low-density polyethylene material is chemically compatible with Reactor Coolant System (RCS) components. It is also chemically compatible with steam generator secondary side chemistry control and components within the secondary cycle. Trace

contaminants in the material have been satisfactorily reviewed within the site consumables program for this use in the RCS system and the secondary system.

The polyethylene mechanically disintegrates by melting below the normal operating temperatures of the reactor coolant. Prior to melting, the mechanical profile is

essentially the same as that of the previously reviewed stainless steel.


any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), was not increased as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the

UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-5

S023-V-1.15. Revision 0:

Title: Reactivity Measurement and Analysis System (RMAS)

Description:

Revision 0 of S023-V-1.15 was developed to setup and use the Reactivity Measurement and Analysis System (RMAS) for Low Power Physics Testing (LPPT).

Safety Evaluation:

This new procedure, S023-V-1.15, approved the use of a new Reactimeter and associated analysis software during Low Power Physics Testing (LPPT). The complete system is referred to as the Reactivity Measurement and Analysis System (RMAS). RMAS was purchased from Framatome Technologies who developed, assembled, and tested the complete system according to their Quality Assurance program. This system provides real time data importing and analysis. It provides superior reliability and redundancy versus the previous system.

RMAS includes a Reactimeter and several workstations (Personal Computers) which communicate with the Reactimeter via serial cable connections. The workstations run the analysis programs which utilize data from the Reactimeter. The actual physics tests (i.e., the parameters measured) are no different with this system versus the previously approved system. The primary calculated input used for physics testing is reactivity. Reactivity is determined from inputs supplied via uncompensated ion chambers (UICs) and predicted kinetics constants (lambda's and beta's). RMAS is the same as the previously approved system in this regard. The primary difference between RMAS and the previous system is the capability to analyze and determine results using software via real time data importing versus manual analysis of data plotted from a strip chart recorder. Although using RMAS to analyze data versus using manual analysis is likely to produce small differences in results, RMAS results are superior in terms of repeatability. In addition, hardware utilized and associated calibration requirements for RMAS do not significantly change .the accuracy of the results versus the previously approved system.

The objectives of the Low Power Physics Test are (1) validating the adequacy of physics models employed in core design performance and safety analyses, (2) verifying compliance with Technical Specification shutdown margin requirements, (3) ensuring proper fuel fabrication and reassembly of the reactor core, including CEA coupling, and (4) assuring CEA integrity. These objectives are not changed as a result of using RMAS.

The tests performed during Low Power Physics Testing are specified by S023-V-1,

5-6

Low Power Physics Testing. In addition, S023-V-1 specifies the acceptance criteria for each test. The use of RMAS is compliant with requirements of S023-V-1 in terms of parameters measured, methodology, and acceptance criteria.

Finally, the use of RMAS does not create any plant condition, manipulation, or equipment interface which is materially different from the previous system.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), was not increased as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-7

S023-2-11 Revision 11:

Title: Turbine Plant Cooling Water (TPCW) System Operation

Description:

Single pump operation for the TPCW system is allowed when one unit is shutdown vice 2 pump operation.

Safety Evaluation:

This change allowed the plant to run in a one pump - one heat exchanger alignment, providing cooling water to both units, during a refueling outage only. Per UFSAR Revision 13, Section 9.2.8.1, the TPCW system has no safety design basis. The TPCW system is not credited with initiating or mitigating any accidents or malfunctions. Reducing cavitation of the pumps may decrease the probability of malfunction.

When two pumps are running with one unit shut down, each pump is typically pumping less than 9000 gpm. The ideal operating point for the TPCW pumps is approx 18000 gpm. Operating with one pump allows the pump to operate much closer to this setpoint (approx 16000 gpm). Operating at less than 9000 gpm in the two pump mode results in severe cavitation to the pumps, which increases the probability of damage to the pump and other components. Operating with one pump reduced the risk of such damage.


5-8

S0123-1-2.5, Revisions 5 and 6:

Title: Battery Rapid Recharge Following a Service Test

Description:

A Rapid Recharge method was used to restore the 1 E batteries to Operable following a Service Test.

Safety Evaluation:

The procedure deals only with the battery maintenance activities. A Rapid Recharge method reduced the time previously spent to fully charge the battery. This eliminated 3 days of float charging. During the Rapid Recharging process, 110% Amp-hour is returned to the battery. A stabilized float charging current of less than 2 amps was verified at the end of recharging. Technical Specification Table 3.8.6-1 Category C limits must be met for the pilot cells as a minimum requirement to declare the battery operable. The battery service test and recharging is performed on an inoperable battery during a refueling outage. The battery is disconnected from the operable plant loads during the testing, so there is no possibility of any plant impact from the maintenance activity. There is also no impact on any accident mitigation functions.

The rapid recharge technique has been approved by both the battery manufacturer and the IEEE and the in-house testing has confirmed that the technique is acceptable to restore battery operability. Battery parameters are closely monitored to ensure that no design limits are exceeded. As such, there has been no damage or loss of reliability for the battery.


5-9

S023-1-8.7. Revision 6:

Title: Emergency Diesel Generator (EDG) Preventive Maintenance Program

Description:

A condition based maintenance strategy for the EDGs is being used instead of the vendor recommended scheduled maintenance.

Safety Evaluation:

The condition based maintenance strategy relies upon a comprehensive engine monitoring program that utilizes current and historical data to assess the health of the engine and it's various components. The premise of this approach is that component degradation will be reflected in various performance characteristics of the engine and can be monitored. Conversely, any wear or degradation that does not affect these performance characteristics is acceptable. The engine monitoring program provides specific requirements that are used to periodically assess the condition of engine components, before any degradation can affect engine Operability.

The design, function, and operation of the EDGs remain unchanged and the principal assumptions and inputs considered in the Updated Final Safety Analysis Report (UFSAR) chapter 15 analysis were not affected by this proposed change and remain valid. This procedure change did not affect the failure modes and effects analysis for any of the EDG support systems as described in the UFSAR. EDG availability and reliability was improved through the reduction of unnecessary intrusive maintenance activities.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the UFSAR, was not increased as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-10

S0123-111-5.5.23. TCN 12-1

Title: Units 2/3 GA Airborne Radioactivity Monitor Sample

Description:

Radiation Monitor 2(3)-RT7870 Sample flow range was increased.

Safety Evaluation:

Radiation monitor 2(3)-RT7870 operating range was increased to allow wider process (effluent) flow range greater operational flexibility in response to air in-leakage conditions. This change reflects testing of the sampling system which demonstrated its ability to maintain isokinetic sampling under a wider range of process (effluent) flow conditions, ensuring compliance with regulatory requirements. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This range increase has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-11

S023-13-13. Revision 6

Title: Misaligned or Immovable Control Element Assembly (CEA)

Description:

Allows Reactor Coolant System (RCS) dilution, allows insertion of CEAs, and incorporates format, administrative, and clarification changes.

Safety Evaluation:

Allowing RCS dilution minimizes the power reduction following a misaligned/dropped CEA and allowing limited CEA insertion if the plant is in Mode 2 near critical provides for more efficient plant operation while preserving plant shut down margin. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This revision has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-12

S023-3-1.1, Revision 21:

Title: Reactor Startup, Administrative Change

Description:

A method was provided to determine the dilution required when the reactor remains sub-critical with all Control Element Assemblies (CEAs) fully withdrawn [All Rods Out (ARO)] during an initial approach to criticality.

Safety Evaluation:

The reason for this change was to allow Reactor Engineering the ability to adjust the targeted Estimated Critical Position (ECP) for the second approach to critical, accounting for the knowledge gained during the first approach to critical. This change does not change the manner in which the dilution is performed, nor does it change the subsequent approach to critical. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This revision has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-13

S023-6-35, Revision 2

Title: Heat Trace Operation

Description:

Alignment attachments for all Heat Trace Panels and breaker number with circuit supplied cross references were created.

Safety Evaluation:

All affected systems meet the design, material, and construction standards applicable to the system or equipment in that system. The probability of occurrence or the

consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this procedure change. This revision has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-14

S023-1-4.16, TCN 2-2:

Title: Pressurizer Heater Inspection and Replacement

Description:

All pressurizer heater acceptance criteria were deleted regarding the phase to ground megger check of >200 megohm and were replaced with a criterion of > 1.5 megohms. Also, the acceptance criteria for phase to phase continuity checks at the heater and at the cubicle were corrected.

Safety Evaluation:

The reason for lowering the megger acceptance criteria is to remove the overly conservative value of 200 megohms, which is the manufacturing acceptance criteria. A megger acceptance value of > 1.5 megohms is acceptable sinc.e a megger value of between 1 megohm and 200 megohms does not affect the heater design life; the intent of the megger verifies that no continuity (shorting or grounding) exists between the terminals or between the terminals and ground. Further, heater phase coils typically fail as a result of the heaters being cycled on and off, similar to how a lightbulb fails (high inrush current) and is not dependent on the results of the megger test. Also, the reason for revising the continuity acceptance criteria is to aid in identifying which pressurizer heaters have degraded performance since resistance readings above 9.0 ohms do not necessarily indicate that a heater has failed. Consequently, it was determined that the phase-to-phase resistance readings at the cubicle should not exceed 10.2 ohms while the phase-to-phase resistance readings at the pressurizer heaters should not exceed 10 ohms. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-15

S023-11-9.712, TCN 3-1

Title: Steam Generator (SG) E088 and E089 Surveillance Testing

Description:

The SG Blowdown Bypass Flow Channel Functional Test was revised.

Safety Evaluation:

The test signal will be injected farther downstream in the circuit to eliminate the need to secure blowdown. A review of the maintenance history of the transmitters has shown that they are highly reliable. The improvement in SG chemistry performance attained by eliminating these 92 day channel functional test's removal of blowdown from service reduces the probability of SG degradation and the likelihood of SG tube failure. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This TCN has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-16

S023-V-1.27, Revision 0

Title: Calculation of Linear Power Subchannel (LPS) Calibration Currents

Description:

Instructions are provided for calculating LPS calibration currents (LPSCC) to correctly determine and document design input values, and to calculate conservative settings for the linear power calibration potentiometer at startup and after fuel reload.

Safety Evaluation:

This new procedure delineates the step-by-step process to calculate new LPSCCs. This procedure was initiated since it is fairy routine (performed every startup) and directly affects the Plant Protection System (PPS) operation and operability. Use of this new procedure decreases the chances of incurring a design input or logic error which could result in erroneously non-conservative LPSCCs The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report

(UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This revision has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-17

S023-V-2, Revision 11:

Title: Power Ascension Testing

Description:

A methodology was added to change Linear Power Subchannel Gains (LPSGs) at the 80+ 5% power plateau, at full power, or after S023-V-2 is closed out.

Safety Evaluation:

The LPSG adjustments, followed by an excore calibration, assure proper performance of the excore detector system with respect to axial power distribution synthesis and high linear power setpoints, including linear power inputs to Core Protection Computer (CPC) - generated trips. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This methodology has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-18

S023-V-3.13. Revision 15

Title: Containment Penetration Leak Rate Testing

Description:

Administrative Leakage Limits were revised for Local Leak Rate Test (LLRT) Components

Safety Evaluation:

In the past, administrative leakage limit revisions have been done on an as-needed basis, usually in response to an LLRT test that failed to meet the limit. This revision is the first comprehensive effort in reviewing and revising all appropriate administrative limits to minimize the need for future administrative limit revisions. These procedure revisions are consistent with procedure guidance and industry practice, are well within Technical Specification limits, are in compliance with the Maintenance Rule program, and ensure continuing favorable comparison with the industry actual LLRT leakage performance. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This revision has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-19

S023-XIII-44, Revision 5

Title: 36-Month Fire Hose Station Functional Test

Description:

The acceptance criterion for fire hose length was changed from 75 ft to 100 ft to ensure hose access and capability through manual fire fighting. Also, the hose service test acceptance criterion was changed from 300 psi to 250 psi, in accordance with the hose's stenciled service test value of 250 psi.

Safety Evaluation:

The provision of 100 ft of hose ensures that any location where safe shut down equipment is required can be reached by manual fire fighting capability. The service test pressure change is to assure that the fire hose tests in the field do not subject the hose to its proof test pressure. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. These changes have no effect on either the existing Limiting Conditions for Operation or the Surveillance requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-20

S023-XIII-57, Revision 8

Title: Barrier Inspections

Description:

Reduction of inspection scope for inspection of internal fire seals. Additionally, wall, floor, and ceiling inspections were combined with penetration seal inspections.

Safety Evaluation:

The periodic inspection of internal fire seals in conduits less than 3 inch diameter, and in conduits up to 4 inch diameter if the conduit lineal run is more than 10 feet from the penetration, was alleviated. This reduction in inspection scope is supported by the San Onofre Nuclear Generating Station (SONGS) Fire Boundary Penetration Seal Evaluation Program Final Report (Doc # 89032) dated July 7,1988 which was approved by the NRC in a Supplemental Safety Evaluation Report (SER) on October 27, 1989. The submitted report contained tests performed to demonstrate that no internal seal is required for conduits 4 inch or less diameter and extending less than 10 feet from the barrier/wall. Combining wall, floor, and ceiling inspections with penetration seal inspections provided a more efficient inspection practice without compromising the quality of inspections. This change does not alter/or lessen the design requirements (fire rating) of the fire area barriers as delineated in the Licensee Controlled Specifications (LCS) 3.7.104, Fire Rated Assemblies. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This combination of inspections has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5-21

S023-XXVI-1 1.6926.1.1 , Revision 0:

Title: Special Test for Relocation of Rad Monitor Power Supply 2/3R17825-2, Rad Monitor 2/3 R17825-2, Control Room Isolation System (CRIS) and Fuel Handling Isolation Signal (FHIS) Functions to the 'TEMP" Panel

Description:

This is a new procedure to support plant continued operation during construction of plant radiation monitor modifications.

Safety Evaluation:

This procedure change was evaluated to ensure compliance. The procedure is to ensure that Operations appropriately aligns all of the affected and associated circuits during this temporary configuration during radiation monitor modifications. Manual actuation to initiate the Control Room Emergency Air Clean Up System (CREACUS) during the time of transfer to or from the temporary panel was consistent with NRC Information Notice 97-78. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This power supply relocation has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the

margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

5059pro

5-22

SECTION 6


UNITS 2 AND 3


Licensee Controlled Specifications (LCS) Changes

6-1

Licensee Controlled Specification (LCS) Change L98-012

LCS Change L98-012 revised LCS 3.3.104.4, "Seismic Monitoring Instrumentation." This change allows a 25% extension to be applied to the 24 month surveillance frequency for a CHANNEL CALIBRATION of Seismic Monitoring Instrumentation (SMI). There was no change to the 24 month surveillance frequency.

Safety Evaluation:

There was no change to the design or operation of the SMI. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-2


LCS Change L98-019 revised LCS Table 3.8.100-1, "Containment Penetration Conductor Overcurrent Protective Devices." This change added 4 molded case circuit breakers to the table which had been erroneously omitted from the LCS.

Safety Evaluation:

There was no change to the design or operation of the circuit breakers. There are no accidents which are initiated by failure of one or more containment electrical penetration overcurrent devices. This change updated the LCS to be consistent with design documents. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-3


LCS Change L99-011 created LCS 3.8.100.2, "Diesel Generator Unit to Unit Cross-Tie, 1 OCFR50.54(x) Permissive Hand Switches." This new LCS provided surveillance requirements for hand switches which provide assurance that the Diesel Generator Cross-Tie would be isolated from the 4.16 kV electrical systems during design basis operations. The Diesel Generator Cross-Tie (including the hand switches) was installed by a design change package evaluated under a 1 OCFR50.59 which was included in the Cycle 10 Facility Change Report.

Safety Evaluation:

There was no change to the design or operation of the Diesel Generator Cross-Tie, the hand-switches, or the 4.16 kV electrical system. This LCS change merely provides assurance that, consistent with its design, the Diesel Generator Cross-Tie will be isolated from the 4.16 kV electrical system during design basis operations. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-4


LCS Change L99-013 revised LCS 3.1.106, "CEA Position Indicator Channels." This change deleted an alternative requirement to shut the reactor down to MODE 3 when a CEA position indication channel is inoperable in MODE 1 or 2.

Safety Evaluation:

CEA position indication channels provide information on the position of CEAs, and are not accident initiators. CEA position indicators are not relied on for adequate reactivity control, they do not provide an indication of a degradation of the reactor coolant pressure boundary or an indication of a challenge to the integrity of a fission product barrier. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-5


LCS Change L99-014 revised LCS 3.1.110, "Remote Shutdown Monitoring Instrumentation - Appendix R Events." This change deleted an alternative requirement to shut the reactor down to MODE 3 when a Control Element Assembly (CEA) position indication channel is inoperable in MODE 1 or 2.

Safety Evaluation:

The Appendix R Remote Shutdown Monitoring Instrumentation is not an accident initiator. Plant safety analyses do not require this instrumentation. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-6


LCS Change L99-018 revised LCS 3.4.101, "Chemistry." This LCS change revised the required actions when a chemistry parameter exceeded its transient limit. Following this LCS change, a return to power was allowed if all chemistry parameters were restored to within the transient limit prior to reaching MODE 5.

Safety Evaluation:

Following this change, if a chemistry parameter exceeds its transient limit, the required action is to begin a reactor shutdown, ultimately reaching Mode 5 within 36 hours. However, if chemistry parameters are restored to within their transient limits prior to reaching Mode 5, the potential no longer exists for RCS chemistry to degrade the RCS boundary. At this point, it would no longer be necessary to continue shutdown. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-7


LCS Change L99-019 revised LCS 3.7.105, "Fire Suppression Water Systems." The LCS was revised to clarify what constitutes an inoperable flow path, and provided new required actions for inoperable flow paths. The new required actions included an increased Allowed Outage Time (AOT) for a firewater pump, with compensatory actions provided during the extended AOT.

Safety Evaluation:

No physical changes were made to the fire protection system's operation or components which would change its performance from the original design intent. Radiological consequences of a fire remain unchanged. No new failure modes were created. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-8


LCS Change L99-021 revised multiple LCSs to provide a default action when the required actions and/or completion times of the various LCSs could not be met. This default action was to "Perform a Cause Evaluation." This new default action replaced a

previous default action to provide a "Special Report" to the NRC.

Safety Evaluation:

Changing the default action has no effect on the design or operation of the systems, structures, and components described in the various LCSs affected by this change. There were no identified commitments regarding Special Reports which were affected by this change. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than

previously evaluated in the UFSAR was not created as a result of this change. There

was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-9


LCS Change L99-023 revised LCS 3.7.113, "10CFR50 Appendix R Safe Shutdown Components." This LCS change removed unnecessary fire zones for fire watch patrols for Train B Salt Water Cooling (SWC) pumps where there is no cable routed for these pumps. This LCS change also added fire zones for Train A SWC pumps where the power cable of one Train A SWC pump is not fully fire protected.

Safety Evaluation:

This LCS change did not adversely impact Appendix R Safe Shutdown capability, and did not impact any systems and equipment administrative controls for any other analyzed accidents and events. There was no modification of any system, structure, or component. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-10


LCS Change L99-025 revised LCS 3.7.103. This change was a clarification that firewatch patrols used as a compensatory action for inoperable spray/sprinkler systems are not required in those fire/area zones inside containment. Also, this LCS change clarified that only one OPERABLE battery bank is required to consider the diesel fire pump OPERABLE.

Safety Evaluation:

This change was a clarification only. The change did not conflict with or invalidate any previous fire hazards analyses nor impact the events and accident analyses bases. The National Fire Protection Association (NFPA) code requires two battery banks for the diesel-driven fire suppression water pump, where one is an installed spare. Only one battery bank is required to fulfill the design requirements to start the fire suppression water pump. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-11


LCS Change L99-026 revised LCSs 3.3.110, "Remote Shutdown Monitoring Instrumentation (RSMI) - Appendix R Events," and 3.7.113, "10 CFR 50 Appendix R Safe Shutdown Components. This LCS change made several changes, listed below:

1. Deleted LCS 3.3.110 in its entirety and moved it to LCS 3.7.113.3. 2. Added a required action for providing compensatory measures within 7 days to

the new LCS 3.7.113.3 3. Identified specific operating Mode applicability for each component in LCS

3.7.113.1 and 3.7.113.3 4. Eliminated duplicate LCS action requirements from LCS 3.7.113.1 for

instruments 2(3)TI-01 11 BY and 2(3)JI-0005C2 and their specific surveillance requirements, because these instruments are covered by LCS 3.7.113.3.

5. Added S2(3)1505MA370 and 2(3)XS41 1, not previously in Table 3.7.113-1. 6. Added new Surveillance Requirements 3.7.113.18, 3.7.113.19, and 3.7.113.20,

to verify operability of the Salt Water Cooling Pump HVAC Fans, Charging Pump Room AC Units, and Control Room Essential Lighting Transfer Switch(es).

7. Removed specific locations of compensatory measures listed in Tables 3.7.1131 and 3.7.113-2, and referencing site controlling procedure instead.

8. Deleted current Table 3.7.113-3, which listed Fire Areas/Zones for implementing compensatory measures. These Tables are also listed in the site controlling procedure.

9. Performed editorial changes to clarify information presented in Table 3.7.113-1.

Safety Evaluation:

This change did not involve physical modification of any structures, systems or components. In addition, the LCS change did not alter the equipment administrative controls for any other analyzed accidents and events. The changes did not reduce any administrative protection provided for Appendix R Safe and Alternative Shutdown equipment. In fact, the changes provided additional protection by adding a new requirement to implement compensatory measures for inoperable Alternative Shutdown instruments. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-12

Licensee Controlled Specification (LCS) Change LOO-002

LCS Change LOO-002 revised LCS 3.2.100, "Linear Heat Rate (LHR)." This LCS change lowered the LHR Power Operating Limit (POL) in order to increase the available large break loss of coolant accident (LBLOCA) peak clad temperature margin from 3 degrees F to approximately 60 degrees F.

Safety Evaluation:

LCS change LOO-002 revised the Linear Heat Rate (LHR) Limit in LCS 3.2.100 from 13 kw/ft to 12.8 kw/ft. The LHR was changed because reducing the LHR thermal power operating limit (POL) results in an increase in the available large break loss of coolant accident (LBLOCA) peak clad temperature margin from 3 F to approximately 60 F. This LCS value establishes the initial conditions at the start of an of accident. It does not contribute to any type of failure that could impact the probability or likelihood of occurrence of any accident described in the Updated Final Safety Analysis Report (UFSAR). The only UFSAR Chapter 15 accident that directly credits the LCS LHR limit is the LBLOCA. Reduction of the LCS LHR limit trades off reduced LHR thermal POL for LBLOCA safety analysis margin. Lowering the LCS LHR limit restricts the allowable plant initial conditions such that the peak clad temperature consequences of the LBLOCA analysis will decrease and consequently be further below the acceptance criteria for this event. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the UFSAR, did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-13


This LCS change re-located the details of the Containment Tendon Surveillance Program from LCS 3.6.100, "Pre-stressed Concrete Containment Tendon Surveillance Program," to the Inservice Inspection Program. As part of this change, three of the acceptance criteria for the Tendon Surveillance Program were revised to be consistent with the current version of 1 OCFR50.55a. The three acceptance criteria that were revised are 1) Tendon elongation, 2) grease void ratio, and 3) Tendon lift-off forces. Tendon elongation and grease void ratio acceptance criteria were relaxed from 5% to 10%. The tendon lift-off force was revised to be consistent with Section IWL of the ASME code.

Safety Evaluation:

There was no change to the design or operation of the Containment structure or tendons as a result of this change. 1 OCFR50.55a, revised in 1996, provides for the 10% acceptance criteria for tendon elongation and grease void ratio. 1 OCFR50.55a specifies that the IWL section of the ASME Code be used for evaluating Tendon Lift-off forces. Changing the acceptance criteria for tendon elongation, grease void ratio, and tendon lift-off force was necessary to be consistent with the 1996 rule change. Therefore, implementing the Tendon Surveillance Program in accordance with this change ensured that the containment tendons and structure would continue to perform as designed. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

6-14


The proposed change revised LCS 3.3.102, "Radiation Monitoring Instrumentation (RMI)," to clarify that inoperability of the alarm or trend function of an RMI channel does not make the entire channel inoperable, provided that adequate compensatory measures are implemented. This may result in plant operation of indefinite duration with an LCS 3.3.102 RMI channel having no Control Room alarm or trend function. The activity was necessary to provide more flexibility of operation by maintaining an RMI channel OPERABLE when it continues to provide valid indication of radiation level and any associated automatic action remains OPERABLE.

Safety Evaluation:

Compensatory measures for the lack of the LCS 3.3.102 RMI channel alarm or trend function provide equivalent protection to that afforded by the alarm or trend function. Moreover, malfunction of the LCS 3.3.102 RMI is not evaluated in the Updated Final Safety Analysis Report (UFSAR), and the LCS 3.3.102 RMI do not appear in the Technical Specifications. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the UFSAR, did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases of any Technical Specification as a result of this change.

50591cs

6-15

SECTION 7


UNITS 2 AND 3

FACILITY CHANGES IMPLEMENTED FOR THE PERIOD FROM MAY 9, 1999 THROUGH FEBRUARY 3,2001

Technical Specification (TS) Bases Changes

"7-1

TS Bases Change B97-040

TS Bases Change B97-040 revised the Basis for LCO 3.3.7, "DG-Undervoltage Start." The change clarified that the CHANNEL CALIBRATION required by Surveillance Requirement (SR) 3.3.7.3 and the response time test required by SR 3.3.7.4 include the entire channel. In addition, TS Bases Change clarified that the time delay for the Loss of Voltage function applies to a step change from nominal (120 V) to zero volts, and changed the term "DG-LOVS" (Diesel Generator - Loss of Voltage Signal) to "Degraded Voltage and Loss of Voltage."

Safety Evaluation:

There is no negative change to the ability of the LOVS to mitigate accidents. Upon loss of all normal alternating current (ac) power the plant design features will initiate automatic startup of the standby diesel generators, which provide adequate power output sufficient to supply electrical power to all necessary engineered feature's systems. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases for the Technical Specifications as a result of this change.

7-2


TS Bases Change B99-014 revised the Basis for LCO 3.7.6.1, "CST T-1 21 and T-1 20." The change deleted a sentence which described the required volume of cooling water (144,000 gallons) in Condensate Storage Tank (CST) T-121 as "corresponding to 96% level." This change was made because the accuracy of the level instrumentation (+/4%) was such that it could not be used as a valid method of determining adequate tank volume. The Surveillance Requirement (SR) may be satisfied by observing the CST T121 low level annunciator.

Safety Evaluation:

The required cooling water volume of CST T-1 21 was not affected by this change. The required cooling water volume continued to be verified following this change by observation of the CST T-121 low level annunciator. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases for the Technical Specifications as a result of this change.

7-3


TS Bases Change B99-016 revised the Basis for LCO 3.8.1, "AC Sources - Operating." The change provided guidance for identifying "required feature(s)" to be declared inoperable when one diesel generator is inoperable and the redundant required feature(s) become(s) inoperable, in accordance with Required Action 3.8.1 .B.2.

Safety Evaluation:

This change provided clarifying information only. There was no change to the Required Action(s) for the case of an inoperable diesel generator. There was no change to the "required features" which must be declared inoperable in accordance with Required Action 3.8.1 .B.2. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases for the Technical Specifications as a result of this change.

7-4


TS Bases Change B99-017 revised the Basis for LCO 3.4.9, "Pressurizer," to implement a reduction in the water level for pressurizer operability from approximately 60% to less than or equal to 57%. These changes were reviewed and approved by the NRC as License Amendments 155 and 146 for units 2 and 3, respectively.

Safety Evaluation:

This change implemented license amendments that had been reviewed and approved by the NRC. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases for the Technical Specifications as a result of this change.

7-5


TS Bases Change B99-018 revised the Basis for LCO 3.4.10, "Pressurizer Safety Valves," to implement an increase in the Pressurizer Safety Valve as-found setpoint tolerance from +/- 1% to +3%/-2%. These changes were reviewed and approved by the NRC as License Amendments 156 and 147 for Units 2 and 3, respectively.

Safety Evaluation:

This change implemented license amendments that had been reviewed and approved by the NRC. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases for the Technical Specifications as a result of this change.

7-6


TS Bases Change B99-020 revised the Bases for LCO 3.4.14, "RCS PIV Leakage." This change package provided a clarification of the amount of leakage which constitutes "flow through the valve" such that a surveillance must be performed to determine whether RCS leakage is within limits.

Safety Evaluation:

This change only quantifies and clarifies the amount of flow considered to be sufficient to disturb the Pressure Isolation Valves (PIVs). This quantification of forward flow required to initiate seat leakage testing will not have any negative impact on the performance of the PIVs. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases for the Technical Specifications as a result of this change.

7-7

TS Bases Change BOO-005

TS Bases Change BOO-005 revised the Basis for LCO 3.1.10, "Boration Systems Shutdown." The change clarified that the minimum required Refueling Water Storage Tank (RWST) level depends on whether the two tanks of a Unit are cross connected (15.5% minimum level) or isolated (17.0% minimum level).

Safety Evaluation:

The proposed change imposes a conservative lower bound on RWST level when the two tanks are isolated. This ensures that the volume of dissolved boric acid assumed in the accident analyses is present for accident mitigation. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases for the Technical Specifications as a result of this change.

7-8


TS Bases Change BOO-007 revised the Basis for LCO 3.4.3, "RCS Pressure and Temperature Limits." The change clarified that a Reactor Coolant System (RCS) temperature change of less than 10 degrees F in any one hour does not constitute a heatup or cooldown during normal operation.

Safety Evaluation:

This change clarified the heatup/cooldown operations noted in TS 3.4.3.1 for consistency with existing TS 5.5.2.4. TS 5.5.2.4 provides controls for tracking RCS heatup/cooldown cycles to ensure the heatup/cooldown cycles are within the component's total number of heatup/cooldown cycles as allowed per the ASME III Code Class 1 stress analysis and within the design limits.

Any change in RCS temperature of less than 10 degrees F in any one hour period during normal operations is considered to be an insignificant transient in the contribution to the component fatigue usage factor as determined by the ASME III Code Class 1 stress analysis. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases for the Technical Specifications as a result of this change.

7-9


TS Bases Change BOO-024 revised the Basis for LCO 3.6.8, "Containment Dome Air Circulators (CDACs)." Technical Specification Basis B 3.6.8 previously described that the airflow from both Containment DACs in a train are required to promote a uniformly mixed containment atmosphere following an accident. This TS Bases change redefined the OPERABILITY requirement for a train of CDAC as one circulation fan, and motor, and the associated controls. This, in turn, re-defined the containment postaccident mixing requirement from 74,000 cfm to 37,000 cfm. No actual physical changes to the dome air circulators were made.

Safety Evaluation:

The dome air circulator system is redundant to the Containment Spray System (CSS) for mixing the containment dome region during a LOCA to prevent formation of hydrogen and radioiodine pockets. This proposed change revised the Technical Specification Basis to require only one dome air circulator per train. CCN N-1 to Calculation M-0072-01 0, "Containment Bldg. Dome Air Circulating System CFM," stated that crediting only one circulator would reduce the dome air turnover rate from 5 to 2.5 per hour and indicates that having one DAC per train is sufficient to help provide adequate mixing of the dome air region. Consequently, the Technical Specification Basis change did not adversely impact the hydrogen mixing function post-LOCA. Therefore, the probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. There was no reduction in the margin of safety as defined in the bases for the Technical Specifications as a result of this change.

7-10

SECTION 8


UNITS 2 AND 3


Non-Conformance Report (NCR) Safety Evaluations

8-1

Unit 2 and 3 NCR 00401086

Title: Piping Insulation in the Emergency Core Cooling System (ECCS) Pump Rooms

Description:

Temporary insulation was added to process piping in the Safety Equipment Building pump rooms 002 and 005 to correct inadequate insulation which was not as specified in design documents.

Safety Evaluation:

Heat load and cooling equipment evaluations were performed to determine the resultant room temperatures in the as-found condition under postulated accident conditions. Environmental Qualification (EQ) evaluations determined temperatures in the most limiting pump room were below the EQ limit for short and long term equipment exposure. Temporary piping insulation was added for personnel protection and to provide extra design margin pending permanent insulation installation. The increase in the transient room temperature of the ECCS pump rooms under postulated accident conditions did not adversely affect the ability of the ECCS equipment to operate. The addition of temporary piping insulation did not change the seismic or environmental qualification of any equipment in the pump room. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), will not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. Adding this insulation has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-2

Unit 2 and 3 NCR 001000900

Title: Safety Injection Tank (SIT) Nitrogen Supply Valves - In Service Test (IST) Program

Description:

The IST Program was revised to add the SIT nitrogen supply valves to the IST program and sets an aggregate SIT leakage criteria of 15 psi in 30 minutes at 615 psi for the nitrogen supply and vent SIT boundary valves.

Safety Evaluation:

The change established an integrated SIT leakage criteria for the IST program. The use of the aggregate leak testing criteria meets the accident analyses and is consistent with the small break loss of coolant accident (SBLOCA) analysis methodology changes implemented per license amendments 163 (Unit 2) and 154 (Unit 3).

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-3

Unit 3 NCR 001001109

Title: ME089 Steam Generator Primary Manway Gasket Seating Surface

Description:

The gasket seating surface of the Unit 2 ME089 steam generator primary manway showed evidence of minor mechanical imperfections (i.e., nicks, dings, scratches).

Safety Evaluation:

The evaluated conditions were normal "wear and tear" expected after twenty years of use and good workmanship. The conditions were within normal expectations for the design of these threaded fastener/gasket mechanical joints. The integrity of the S21301 ME089P steam generator primary manway gasket seating surfaces have been evaluated through visual observation and ASME Section Xl Code required examinations performed at the conclusion of the Unit 2 Cycle 10 refueling outage and examinations performed at the beginning of the current Unit 2 Cycle 11 outage. The conditions described in this NCR are minor in nature. They are caused by normal maintenance activities performed on the generators at each refueling outage. Through compliance with ASME Code Section Xl required examinations it has been shown that the current condition of the steam generator primary manway gasket seating surfaces does not constitute an unreviewed safety question.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This evaluation has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-4

Unit 2 NCR 001002022

Title: Train A Containment Emergency Sump Paint Repair

Description:

Peeling paint coating along an approximately four foot long crack on the east wall in the

Train A Containment Emergency sump was removed and repaired.

Safety Evaluation:

The design function of the Containment Emergency Sump is to provide a flow path for

long term Reactor Core and Containment Cooling Post LOCA. The removal of the

peeling paint provided assurance that the water quality being supplied to the ECCS pumps will be free of debris.




UFSAR was not created as a result of this change. This repair has no effect on either

the existing Limiting Conditions for Operation or the Surveillance Requirements in the

Technical Specifications: thus, the margin of safety as defined in the bases for the

Technical Specifications was not reduced as a result of this change.

8-5

Unit 2 NCRs 001100377 and 001100634

Title: Plug Repair of Sleeved Steam Generator Tubes with Circumferential Flaws

Description:

Some sleeved steam generator tubes in ME088 with circumferential flaws failed to meet the sleeve to tube joint weld fusion criteria or were inaccessible for welding. The tubes were removed from service by plugging.

Safety Evaluation:

The tube sleeve acts as a stabilizer and limits deformation caused by fluid-elastic flow induced vibrations in tubes with circumferential flaws, thus preventing damage to adjacent tubes. The presence of a plug in the tubes eliminates primary coolant leakage out of a tube through the circumferential flaw.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This repair has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-6

Unit 3 NCR 010102507

Title: Emergency Core Cooling System (ECCS) Piping Spring Can Support Reading

Description:

A field inspection of Line $3-SI-003-1 1 (Emergency Containment Sump piping back to the ECCS system in the Safety Equipment Building) revealed an As Found condition of one inch off of the Cold Load (CL) setting for a spring can support. The As Found condition for spring can S3-SI-003-H-003-N was 1-1/8 inch off the Cold Load setting. The design specification has a limit of 1/2 inch for the as found condition for either the Hot Load (HL) or Cold Load setting.

Safety Evaluation:

Pipe stress calculations based on the As Found conditions determined that the stress on the ECCS piping was acceptable and ASME code requirements were met. The scale on the spring can support was adjusted to reflect the correct reading.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This repair has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-7

Unit 3 NCR 980301095

Title: Pressurizer Heater A620 Failure

Description:

Failed Pressurizer Heater A620 was left in place for Unit 3 Cycle 11 operation

Safety Evaluation:

Attempts to remove failed Unit 3 Pressurizer Heater A620 during the Unit 3 Cycle 11

refueling outage were unsuccessful. Degradation of the heater sheath internal to the

pressurizer was noted during the repair process. The heater sleeve penetration had

already been capped in a previous outage. The welded cap met all ASME III and Xl

code requirements and did not compromise Reactor Coolant System (RCS) pressure

boundary integrity. Analysis of the heater sheath degradation was performed to ensure

satisfactory operation until the next refueling outage.





UFSAR was not created as a result of this change. This evaluation has no effect on

either the existing Limiting Conditions for Operation or the Surveillance Requirements

in the Technical Specifications: thus, the margin of safety as defined in the bases for

the Technical Specifications was not reduced as a result of this change.

8-8

Unit 2 and 3 NCR 980901673

Title: Recalculation of the Maximum Post-Loss Of Coolant Accident (LOCA) Flood Level in Containment

Description:

The maximum post-LOCA flood level in containment was previously determined in calculation N-4060-030. A calculation revision lowered the design basis flood level from the previously identified value of 25'-0" to a lower value of 24'-6". The reduction in calculated flood level was necessary to prevent Emergency Air Cooling Unit (ECU) ME401 in either Unit 2 or Unit 3 from becoming incapacitated by post-LOCA flooding of

the vertical air shaft through an air supply register with a base elevation of 24'-6". The other three ECUs in Units 2 and 3 are designed to accommodate flood levels to 25'-0". In addition, all other important to safety equipment in containment required to remain operable for mitigation of LOCA or MSLB events in containment are protected against flooding to an elevation of 25-0".

Safety Evaluation:

The revision of flooding calculation N-4060-030 resulted in a reduction in the containment maximum post-LOCA flood level from the previous value of 25'-0" to a new value of 24'-6". The lower value reduces the severity of flooding. Containment post-accident pressures and temperatures will be unaffected, and radiological consequences of any accidents involving containment post-accident flooding will not be changed.


any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-9

Unit 3 NCRs 981001087 and 981001103

Title: Containment Spray Pump Suction Line Relief Valves 3PSV8157 and 3PSV8158 Spring Ratings

Description:

Incorrect valve springs are installed in 3PSV8157 and 3PSV8158. 3PSV8157 and

3PSV8158 have a design setpoint of 110 psig. The installed springs have a range of

141 to 240 psig. This evaluation is Interim Accept-As-Is disposition that restricts the

use of these valves until the Cycle 12 refueling outage when the valves and/or springs

will be replaced.

Safety Evaluation:

The containment spray system is designed to remove particulate airborne iodine fission

products within the containment atmosphere. The function of this 3PSV8157/8 is to perform overpressure protection for the containment spray pump suction line during the

time period in which the ECCS is in Recirculation Actuation Signal (RAS). It is required

to lift at set pressure and relieve the pressure in the line caused by leakage past the

mini-flow check valves. This check valve is routinely tested for back leakage and is

assumed to have a maximum of 1 gpm leakage back past the seat. The design flow

rate of this valve is 5-6 gallons per minute. The stronger spring installed into this relief

valve limits the ability of the valve to pass the design flow rate. This limited flow rate is

still greater than the 1 gpm flow rate which could be leaking past the check valve seat.

Installation of stronger springs into the relief valves results in the possibility that the

valves will not be able to reach their full relief flow capability, however, the valves will

still relieve at the set pressure, and will pass the flow expected from the pump

discharge check valves (which are tested for back leakage).





UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the



8-10

Unit 2 NCR 990200824

Title: Thread in Studhole No.18 of the Steam Generator E088 Cold Leg Primary Manway

Description:

The internal thread in studhole no. 18 of the Unit 2 steam generator E088 cold leg primary manway is degraded.

Safety Evaluation:

Review of design calculations regarding preload of steam generator and pressurizer manways determined that there was sufficient thread engagement in the studhole to preclude stripping the threads during service. The steam generator primary manway threaded fastener joint remains within the design basis. The design basis calculations are in accordance with the rules of American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, ASME Section X1. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This evaluation has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-11

Unit 3 NCR 990402371

Title: Reactor Coolant Pump Motor 3MM002 Lubricating Oil Orifice and Instrumentation Change

Description:

The lube oil system piping orifices in the Reactor Coolant Pump Motor 3MM002 were replaced with blank orifices which defeat the lube oil low flow alarms for the motor.

Safety Evaluation:

The reactor coolant pump motor has an internal heat exchanger in the oil reservoir to cool the oil, and the Anti Reverse Rotation Device (ARRD) bearing temperature is used to monitor that the ARRD is adequately supplied with oil. This NCR does not affect the operation or design basis of the reactor coolant system and/or reactor coolant pump. It does not alter the design bases assumed in the UFSAR or affect any input to any scenario described in the accident analysis. The plant function of the Reactor Coolant Pump Motor is unchanged, and the design basis is maintained. The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance Requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-12

Unit 2 and 3 NCR 000101035

Title: Train B Salt Water Cooling Pump 2(3) P114 Control Circuitry Modification

Description:

Isolation fuses were added in the control circuit of the Train B Salt Water Cooling

Pumps 2(3) P114 to prevent tripping of the DC control power circuit breaker if a fire

should occur in the Salt Water Cooling Piping tunnel (Fire Zones 2-148E, 2-148F, and

3-148F).

Safety Evaluation:

The modification restores the Train B Salt Water Cooling Pump control circuitry to the

condition described in the UFHA. The operational condition and the design bases of

the Salt Water Cooling System are unchanged.



Analysis Report (UFSAR), will not increase as a result of this change. The possibility

of either an accident or malfunction of a different type than previously evaluated in the

UFSAR was not created as a result of this change. This change has no effect on either

the existing Limiting Conditions for Operation or the Surveillance requirements in the



8-13

Unit 2 NCR 000300208

Title: Spare Salt Water Cooling Pump Lower Outer Column Weld

Description:

A weld was repaired on a spare Salt Water Cooling Pump lower outer column that had corroded below the minimum wall thickness requirements.

Safety Evaluation:

The weld repair to the lower outer column of the spare Salt Water Cooling Pump restores the component to the original design configuration. The functionality of the pump is unchanged.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), will not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-14

Unit 2 NCR 000300400

Title: Salt Water Cooling (SWC) Pump Discharge Check Valve S21413MU012

Description:

Some missing internal pieces of Salt Water Cooling Pump Discharge Check Valve S21413MU012 were Accepted-As-Is. The check valves are dual plate check valves that use spring pressure in addition to reverse flow to shut the check valve. The "fingers" that attach the coil spring to the flapper plate have broken off and are missing.

Consequently, there is no spring pressure to assist this valve in closing. The missing parts were not located in an inspection of the SWC piping and heat exchangers.

Safety Evaluation:

The Safety Evaluation determined that fluid backflow was the major driving force to

close the flapper plate. This valve is tested during quarterly ISTs which ensure the valve will perform its design function without the assistance of spring pressure. The missing parts were determined not to pose a challenge to the operability or design basis of the SWC system.


any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), will not increase as a result of this evaluation. The possibility of either an accident or malfunction, of a different type than previously evaluated in the UFSAR was not created as a result of this change. This evaluation has no effect on either the existing Limiting Conditions for Operation or the Surveillance requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-15

Unit 2 and 3 NCR 000401014

Title: Control Room Emergency Air Condition Unit E418 Transfer Switch Repair

Description:

The replacement of six rivets with steel screws and the glue repair of the trip unit cover was Accepted-As-Is. The trip unit cover was damaged during the implementation of a

field change to disable the automatic trip feature of the transfer switch.

Safety Evaluation:

The structural integrity of the transfer switch was not compromised by the repair. The function of the transfer switch was not changed and the design basis was maintained.


any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), will not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-16

Unit 2 and 3 NCRs 000401454 and 000500354

Title: Salt Water Cooling Pump Discharge Actuator Valve Travel Stop

Description:

A valve travel stop was installed on the Salt Water Cooling Pump discharge valve actuator. The valve travel stop will prevent the pneumatically operated Salt Water

Cooling Pump discharge valve from inadvertently closing on a loss of instrument air or

safety-related backup nitrogen accumulators.

Safety Evaluation:

The addition of valve travel stops on the actuator of the Salt Water Cooling Pump discharge valve ensures the valve will remain in the accident analysis required position (open) following a loss of pneumatic motive force. The functionality and the design basis of the Salt Water Cooling System is not changed.



Analysis Report (UFSAR), will not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the

UFSAR was not created as a result of this change. This change has no effect on either

the existing Limiting Conditions for Operation or the Surveillance requirements in the



8-17

Unit 2 NCR 001100873

Title: Control Element Drive Mechanism (CEDM) No. 61 Vent Valve Seal Weld

Description:

The vent valve for CEDM no. 61 had unacceptable leakage past the valve seat. The housing nut for vent valve for CEDM no. 61 was seal welded shut to eliminate the leakage. The vent valve will be replaced at the next outage of sufficient duration and proper plant conditions.

Safety Evaluation:

The seal weld of the housing nut for CEDM no. 61 vent valve retains the pressure integrity of the RCS pressure boundary. The operation of CEDM no. 61 remains unchanged and is consistent with the CEDM and RCS design bases.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), will not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-18

Unit 3 NCR 010101490

Title: Letdown Heat Exchanger Flow Divider Plate Repair

Description:

The flow divider plates in the Letdown Heat Exchanger were found to be cracked during

inspection. The flow divider plates were replaced with 304L/316L stainless steel plates.

Safety Evaluation:

There was no change to the dimensional layout or the design function of the flow

divider plates in the Letdown Heat Exchanger. The flow divider plates do not provide a

pressure retaining function or provide support to the channel head shell. The material properties of the 304L/316L replacement divider plates (thermal expansion, corrosion resistance, strength, etc.) were determined to be acceptable for use. The design basis

of the Letdown System remains unchanged.


any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), will not increase as a result of this change. The possibility

of either an accident or malfunction of a different type than previously evaluated in the

UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance requirements in the

Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-19

Unit 3 NCR 010101893

Title: Potential Loose Part in Steam Generator ME089 Secondary Side

Description:

A foreign object discovered in the secondary side of steam generator ME089 in the stay cylinder area was left in place. The object was wire-like in appearance approximately 0.006" in diameter and 2" long (visible portion). It did not appear to part of the steam generator internals. Attempts to dislodge/retrieve the object were unsuccessful.

Safety Evaluation:

All the steam generator tubes in the vicinity of the object have been plugged. Potential loose part and steam generator tube interaction scenarios - random impact under influence of flow stream; repeated impact on plugged tubes causing tube collapse (Ginna Scenario); and wedged part with tube wear - were evaluated and determined to be acceptable. The potential for a steam generator tube rupture caused by this loose part was considered to be negligible.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), will not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the

UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-20

Unit 3 NCR 980900672

Title: Swing Train Charging Pump Motor 3P1 91 Transfer Switch Circuit

Breakers

Description:

The circuit breakers contained in transfer switch 3DB041 0 for the swing train Charging

Pump 3P1 91 contained automatic trip devices which were not included in design documents or engineering calculations. Per design, the transfer switch circuit breakers

were to be static switches without automatic protective features. These breakers were

Accepted-As-Is until they could be replaced by non-automatic devices.

Safety Evaluation:

Analysis of the time-current protection curves for the transfer switch circuit breakers

and the load center feeder breaker revealed that the presence of the automatic trip devices in the transfer switch circuit breakers would not adversely affect the operation

of the swing train charging pump. In the thermal region of the time-current protection curves, the load center circuit breaker provides primary protection for the charging pump motor. The instantaneous trip of the transfer switch circuit breakers is high enough to preclude spurious tripping during motor start. The functional operation of

the transfer switch is unchanged and the design bases of the Charging System are maintained.



Analysis Report (UFSAR), will not increase as a result of this evaluation. The

possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this evaluation. This change has no effect on either the existing Limiting Conditions for Operation or the

Surveillance requirements in the Technical Specifications: thus, the margin of safety as

defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-21

Unit 2 and 3 NCRs 980900943 and 980900967

Title: Control Room Complex Emergency AC Unit E418/E419 Transfer Switch Circuit Breakers

Description:

The circuit breakers contained in transfer switch DB0415 and DB0615 for the Control Room Complex Emergency AC Units contained automatic trip devices which were not included in design documents or engineering calculations. Per design, the transfer switch circuit breakers were to be static switches without automatic protective features.

These breakers were Accepted-As-Is until they could be replaced by non-automatic devices.

Safety Evaluation:

Analysis of the time-current protection curves for the transfer switch circuit breakers and the load center feeder breakers revealed that the presence of the automatic trip devices in the transfer switch circuit breakers would not adversely affect the operation of the Control Room Complex Emergency AC Units. In the thermal region of the timecurrent protection curves, the load center circuit breaker provides primary protection for

the Emergency AC Unit. The instantaneous trip of the transfer switch circuit breakers is high enough to preclude spurious tripping during motor start. The functional operation of the transfer switch is unchanged and the design bases of the Control Room Complex Emergency AC System are maintained.


any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), will not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. This change has no effect on either the existing Limiting Conditions for Operation or the Surveillance requirements in the Technical Specifications: thus, the margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-22

Unit 2 NCR 980902074

Title: Electrical Penetration Assemblies

Description:

Fuses were added to the electrical penetration lighting circuit associated with the

containment airlock. Addition of fuses to the electrical penetration circuit ensured that

the electrical penetrations could withstand, without loss of mechanical integrity, the

maximum fault current vs. time conditions that could occur as a result of single random

failures of circuit overload devices. This re-established conformance between the

as-built facility and the UFSAR for containment electrical penetration overcurrent

protection.

Safety Evaluation:

There are no accidents evaluated previously in the safety analysis report which are

initiated by failure of one or more containment electrical penetration overcurrent

devices. The proposed change re-established conformance between the UFSAR

design basis and the "as-built" facility for single failure protection of the specific

containment electrical penetration of concern. Appropriate electrical design, material,

and construction standards were implemented thereby maintaining overall electrical

protection system performance.





UFSAR was not created as a result of this change. The margin of safety as defined in


8-23

Unit 2 and 3 NCR 981000725

Title: Component Cooling Water - Backup Nitrogen Supply (CCW - BNS)

Description:

The plant condition in the area of the Backup Nitrogen Supply (BNS) system for the Component Cooling Water (CCW) surge tanks was Accepted-As-Is based on probabilistic risk assessment. A portion of the BNS system is located outside the safety-related structures and configured such that both trains could be potentially impacted by a tornado generated missile. The existing configuration was, however, acceptable "as-is" based on the calculated probability of a tornado missile affecting both trains of BNS (5.4E-1 O/yr) being lower than the regulatory guideline value of 1 E-7/yr.

Safety Evaluation:

Neither the CCW system nor the BNS system can by itself be a precursor to any accident. Accepting the existing configuration of the BNS system does not introduce any new interaction of this system with any system whose malfunction might result in initiation of an accident. The proposed activity accepted the existing configuration of the BNS system based on the fact that the probability of both trains of this system being impacted by a tornado missile is very low. This probability is lower than that specified in the regulatory guidelines, below which tornado missile protection is not required.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the Updated Final Safety Analysis Report (UFSAR), did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-24

Unit 2 NCR 990100632

Title: Broken Stud Bolt on Panel 2Y0407

Description:

Missing/broken studs on the Vital Bus Panel 2Y04 cabinet were Accepted-As-Is. The 3 studs are part of the pattern of two vertical columns of 8-1/4" studs which hold a facia within the left side of the cabinet.

Safety Evaluation:

The facia's purpose is to limit inappropriate access to the fuses and other components in the cabinet. As such, it performs no structural function. There were sufficient bolts remaining to assure that the facia was held in place so that it performs its function. Operability of the inverters and associated AC electrical equipment was not affected.


8-25

Unit 3 NCR 990401629

Title: High Pressure Safety Injection (HPSI) flow indicator

Description:

A HPSI flow indicator (3FI0341-2) reading 21% low was Accepted-As-Is.

Safety Evaluation:

HPSI flow indication is used for In-service Testing (IST) and accident mitigation. It

does not directly influence any equipment or instructions during normal operation and

therefore cannot lead to or increase the probability of an accident. The low indication of

HPSI flow to one cold leg could have conservatively impacted IST surveillance results, but would not have led to passing equipment which does not meet design requirements. The low flow indication could have had a small effect on HPSI control

during a Loss of Coolant Accident (LOCA); however, the impact would have in general

led to slightly higher flow rates with respect to the minimum requirements. The slightly

higher actual flow rates (-<50 gpm) which might have occurred if indicated flow was

permitted to run at the upper limit (910 gpm) still left ample margin to protect against HPSI run-out damage.







8-26

Unit 2 and 3 NCR 990402096

Title: Containment Construction Vent Fill

Description:

A non-conformance of the containment structure was Accepted-As-Is.. 4000 psi concrete was used to fill three 24" diameter construction vents located in the dome portion of the Unit 2 and Unit 3 containment structures. SONGS calculation and drawings indicate 6000 psi concrete for containment construction.

Safety Evaluation:

SONGS Units 2 and 3 containments are prestressed reinforced concrete structures. For typical reinforced concrete design, 3000 to 4000 psi concrete is used for nonprestressed structures and 5000 to 6000 psi concrete is used for prestressed structures. The reason for using the higher strength concretes for prestressing applications is that they exhibit reduced creep and shrinkage effects which results in less prestressing losses. The use of 4000 psi concrete to fill three 24" diameter constructions vents did not impact this design practice, as these construction vents account for less than 0.1% of the volume of concrete in the containment dome.

Concrete strength does not affect the shielding characteristic of the containment structure; therefore, since the thickness of the concrete in the construction vent plugs are the same as the surrounding structure, there is no change in the radiological consequences of any accidents.

The bases for the containment structure's design, specifically integrity, pressure, and temperature, are not dependent upon the change in compression strength from 6000 psi to 4000 psi for the concrete in the three construction vents.


8-27

Unit 3 NCR 990402292

Title: Containment Airlock Electrical Penetration Assemblies

Description:

Fuses were added to the electrical penetration assembly for the lighting circuit for the emergency airlock.

Safety Evaluation:

There are no accidents evaluated previously in the Updated Final Safety Analysis Report (UFSAR) which are initiated by failure of one or more containment electrical penetration overcurrent devices. Appropriate electrical design, material, and construction standards were implemented, thereby maintaining overall electrical protection system performance. Addition of fuses to the electrical penetration circuit ensured that the electrical penetrations withstand, without loss of mechanical integrity, the maximum fault current vs. time conditions that could occur as a result of single random failures of circuit overload devices. In so doing, the containment mechanical penetration will perform its function consistent with the accident analyses for offsite and control room doses.

The probability of occurrence or the consequences of an accident, or malfunction of any equipment important to safety, previously evaluated in the UFSAR, did not increase as a result of this change. The possibility of either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-28

Unit 2 NCR 990500775

Title: Safety Injection Tank (SIT) Vent Valve Stroke Time

Description:

A SIT Vent Valve was stuck partially open which could have delayed Shutdown Cooling (SDC) entry conditions following a Loss of Coolant Accident.

Safety Evaluation:

SIT vent valve 2HV9345 was opened following maintenance to support venting and was found to vent at a rate of -1.4psi/min (i.e., partially opened). However, subsequent testing confirmed that the valve would eventually fully open in approximately 20 minutes, and finally upon demand with no delay, indicating that the valve had resumed normal operation. SIT tank pressure is required to either be vented or isolated from the RCS by means of the SIT outlet valve prior to initiating SDC. Should the valve stick again, SDC entry conditions could be delayed. However, the vent path provided by the pilot valve is adequate to meet safety analysis assumptions

The non-conformance cannot create a sequence of events that lead to any previously evaluated accident and thus cannot increase the probability of these events. Had the valve stuck in a post-accident scenario, it would have been possible to de-pressurize the SIT through a combination of letdown and venting.


8-29

Unit 3 NCR 990501763

Title: Diesel Fuel Oil Storage Tank Level Transmitter

Description:

The diesel fuel oil storage tank low level cutout switch was bypassed. The associated transmitter had been determined to be faulty. The bypass prevented an erroneous low level signal from starting the fuel oil transfer pumps.

Safety Evaluation:

The diesel fuel oil storage system supports the emergency diesel generators to provide

power during a loss of offsite power. The existence of the jumpers did not increase the likelihood of a failure of the fuel oil transfer system because the transfer pumps continued to operate as designed. The fuel oil transfer system continued to operate

and supply the day tank as required for diesel operability. By installing the jumpers, operators would have been able to operate the transfer pumps whether or not an erroneous low level indication occurred due to the degraded condition of the level transmitter. The potential existed for the pumps to cavitate and be damaged had the main fuel oil tank become empty following an accident; however this was not a concern because this would not have occurred until after seven days of diesel generator operation as required by the Technical Specifications.




either an accident or malfunction of a different type than previously evaluated in the UFSAR was not created as a result of this change. The margin of safety as defined in the bases for the Technical Specifications was not reduced as a result of this change.

8-30

Unit 3 NCR 990600753

Title: Startup Test Channel Calibration

Description:

A calibration error for the Auctioneer circuit for a Startup Test Channel was AcceptedAs-Is. The Auctioneer Circuit is designed to switch the startup test channel from the Log Count and Rate (LCR) board to the log amplifier circuit at a pre-set power level. A required calibration adjustment of 60 mV was not performed on the Auctioneer Circuit.

Safety Evaluation:

An auctioneer voltage of 6.34VDC (the as-left Unit 3 Startup Channel 2 Wide Range auctioneer voltage) would result in a crossover setpoint of 2.2E-2% power. A 60 mV adjustment to the setpoint to 6.40 VDC would result in a crossover setpoint of 2.5E-2% power, or a difference of 3E-3% power.

The difference of 3E-3% power in the crossover setpoint was insignificant, especially considering that criticality is achieved well below that magnitude of power, typically in the 1 E-6 to 1 E-5% power range, and that Units 2 and 3 are not normally operated near the 1 E-2 power range (other than during Low Power Physics Testing, during which flux measurements are obtained using Excore Control Channels). Also, the 60 mV difference was small compared 200 mVDC hysteresis which provides stable switching at the crossover point. Thus, the 60 mV adjustment to the auctioneer setpoint is minor compared to the uncertainty in the crossover setpoint.

Since the function of the Unit 3 Startup Channel 2 was not impacted by a 60 mV difference in the Auctioneer setpoint, the as-left Auctioneer voltage of 6.343 VDC was acceptable as the calibration voltages.


8-31

Unit 2 NCR 990600967

Title: Containment Isolation Valve Identification

Description:

A mis-identification of a containment isolation valve was Accepted-As-Is. The valve appeared on Piping and Instrumentation Diagrams (P&IDs) as a globe valve. In reality the valve is a gate valve.

Safety Evaluation:

The subject valve provides containment isolation during normal and emergency plant operation. The gate valve installed in the field was as capable of performing the isolation function as the globe valve shown on the P&ID.


8-32

Unit 2 NCR 990601284

Title: Component Cooling Water (CCW) Heat Exchanger Saltwater Normal

Outlet Valves

Description:

A compensatory action was taken for a circuit deficiency in the control circuits for the

CCW Heat Exchanger saltwater normal outlet valves. This deficiency could have

resulted in the valves going closed and staying closed in the event of a transfer of their

associated 4 kV bus to the opposite unit via the 4 kV cross-tie. The valves were placed

in the open position with power removed (the safe shutdown position), and the input to

the Bypass/Inoperable Status Alarm System from these valves was disabled.

Safety Evaluation:

The annunciation of the status of the Emergency Safety Features (ESF) systems is not

an accident initiator. Placing the valves in the open position ensured that the valves

would be in their safe position and that the Saltwater Cooling System would continue to

perform its design function. The annunciator system is non-safety-related and does not

perform any Emergency Core Cooling System function.







8-33

Unit 2 NCR 990700343

Title: Repair of Feedwater Bypass Valves

Description:

Leaking Feedwater Bypass Valves were repaired using sealant injection.

Safety Evaluation:

Sealant injection did not impact the feedwater bypass valves from performing their design function. Failure of the air supply/control lines due to incidental contact with the injection rig is not considered credible. The consequences of a Loss of Normal Feedwater Flow, a Feedwater System Pipe Break, and an Asymmetric Steam Generator Transient all bound the potential consequences of a feedwater leak or inadvertent feedwater bypass valve closure that could result from installing and maintaining this design modification.


8-34

Unit 2 and 3 NCR 990701248

Title: Component Cooling Water(CCW) System Operation

Description:

Operation of the CCW system was changed to improve reliability. The changes involved isolating the normal Nuclear Service Water (NSW) from the CCW system such that only the safety-related makeup system could be used for makeup.

Also, procedure changes were implemented which revised the method for verifying Non-Critical Loop (NCL) integrity. The new method involved cracking open the NCL isolation valves, rather than stroking them.

Safety Evaluation:

The NSW makeup to the CCW system, which was previously utilized to manually fill the Surge Tank, was isolated such that it could not be used in Modes 1 - 4, and in Modes 5 and 6 when the CCW system must be operable to support the Shutdown Cooling System. The NSW system is non-safety related, and there is no Technical Specification for Nuclear Service Water or any requirement for the NSW system to be operable for accident mitigation. Instead of NSW, the safety-related makeup is to be used. Use of the safety-related makeup for normal filling of the CCW Surge Tank was permitted prior to this change.

Revising procedures to require cracking open the NCL isolation valve instead of fully stroking it minimized the potential for water hammer and also minimized usage of backup nitrogen inventory.


8-35

Unit 2 NCR 991201143

Title: Core Exit Thermocouple (CET) Inoperable

Description:

A CET which was reading significantly higher than all other CETs in the same quadrant was intentionally "failed low" by installation of a jumper wire.

Safety Evaluation:

The CETs are passive reactor monitoring devices. The primary purpose of the CETs is

as an input to Post-Accident Monitoring Instrumentation (PAMI) to provide information to the control room operators during accident conditions. There are six remaining OPERABLE CETs in the affected quadrant, and only 2 are required to be OPERABLE. The loss of one CET and installation of electrical jumpers for does not challenge the operability of the PAMI system.


8-36

Unit 2 and 3 NCR 991201371

Title: Dose Assessment Computer Units

Description:

The Dose Assessment Computer System (DACS) receiving an input in the wrong units.

The Emergency Response Data System (ERDS) was sending the output for Radiation

Monitor 7808C in micro-curies/cc. The DACS was designed to receive this input cpm.

Safety Evaluation:

The data point for radiation monitor 7808C is used for monitoring releases and is not

the sole source for this information. The information provided is used only to aide

decisions for site emergency classifications/actions and does not affect the probability

or consequences of an accident.






the bases for the Technical Specifications was not reduced as a result of this change

5059ncr

8-37

i southern edisondwight california edisonvice president · 2012. 11. 18. · i southern...

Documents