responds to nrc 980518 ltr re violations noted in insp ... · code section xi, jvw-1100,...

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A CMS Energy Company

June 24, 1998

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, Ml 49043

DOCKET 50-255 - LICENSE DPR-20 - PALISADES PLANT REPLY TO NOTICE OF VIOLATION AND NOTICE OF DEVIATION FROM INSPECTION REPORT 50.,255/98003

NRG Inspection Report 50-255/98003(DRS) dated May 18, 1998, cited six violations. The violations involved procedural violations, failure to properly scope and include certain valves with safety functions in the inservice test (IST) program, and two examples of inadequate design control. In addition, two deviations from previous licensing commitments are cited. The responses to the violations and deviations are included in the attachments to this letter.

Consumers Energy agrees with five of the six violations. Violation 50-255/98003-0?b is · being contested; as described in Attachment 6.

Consumers Energy agrees with one of the deviations, however, deviation 50-255/98003-08 is being contested as described in Attachment 8.

We examined the violations, as written, for potential generic implications regarding procedural adherence issues and found no pattern indicative of a process breakdown. The violations have been determined to be isolated instances, which have pointed to unique weaknesses.

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Our previous response to the 1 OCFR50.54(f) letter, dated Febn.iary-6, 1997 includes- - -------·---- -· actions that will further ensure the adequacy of design control.

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·980630003~ 980624 PDR ADOCK 05000255 G . PDR

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SUMMARY OF COMMITMENTS

This letter contains no new commitments and no revisions to existing commitments.

Palisades "60 Day Response to NRC Inspection Report No. 50-255/97-201, 'Palisades Plant Design Inspection"', dated March 2, 1998, committed to provide the NRC with a status of our progress in completing all actions identified as a result of the design inspection report. This action is to be completed by October 1, 1998. ·

~~ Thomas J. Palmisano Site Vice President

CC Administrator, Region Ill, USNRC Project Manager, NRR, USNRC NRC Resident Inspector - Palisades

Attachments

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ATTACHMENT 1

CONSUMERS ENERGY COMPANY PALISADES PLANT

DOCKET 50-255

REPLY TO NOTICE OF VIOLATION 50-255/98003-03

FAILURE TO PROPERLY SCOPE VALVES CK-ES3339, CK-ES3340, CK-DMW400, CV-1813, CV-1814, CV-1501, CV-1502 AND CV-1503

AND INCLUDE THEM IN THE IST PROGRAM

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ATTACHMENT 1 REPLY TO NOTICE OF VIOLATION 50-255/98003-03



NRC NOTICE OF VIOLATION

During an NRG inspection conducted from March 25, 1998, through April 10, 1998, six violations of NRG requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRG Enforcement Actions," NUREG-1600, violation 50-255198003-03 is listed below:

A. Technical Specification 6.5. 7, "lnservice Inspection and Testing Program," states, in part, that this program provides controls for inservice inspection and testing of ASME Code Class 1, 2, and 3 components including applicable supports.

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI, JVW-1100, "Valve Testing, "states that valve testing shall be performed in accordance with the requirements stated in OM-10. Section 1. 1, "Scope," of OM-10 states the active and passive valves covered are those which are required to perform a specific function in shutting down a reactor to cold shutdown condition, in maintaining the cold shutdown condition; or mitigating the consequences of an accident.

Contrary to the above, the following valves' functions were not included in the inservice test program although they performed a specific function in shutting down a reactor to cold shutdown condition, in maintaining the cold shutdown condition, or mitigating the consequences of an accident.

· 1. As of November 10, 1997, check valves CK-ES3339·and CK-ES3340 in the minimum flow recircula_tion piping from the discharge of each high pressure safety injection pump had a safety function to close to prevent the potential overpressurization of the pumps suction piping and were not included in the /ST Program.

2. As of March 17, 1998, check valve CK-DMW400 in the flow path from the primary system make-up storage tank T-81 to the condensate storage tank T-2 had a safety function to open, supplying make-up to the

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• A TT AC HM ENT 1 REPLY TO NOTICE OF VIOLATION 50-255/98003-03



condensate storage tank. This check valve was not included in the /ST Program.

3. As of March 17, 1998, control valves CV-1813 and 1814 in the containment purge and ventilation system had an active safety function to

. close to provide a containment isolation function and were not included in the /ST Program.

4. As of March 26, 1998, control valves CV-1501, 1502, and 1503 in the . plant heating system had an active safety function to close to provide a containment isolation function and were not included in the /ST Program.

This is a Severity Level IV Violation (Supplement 1). (50-255198003-03)

• CONSUMERS ENERGY RESPONSE

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Consumers Energy Company agrees with this violation as written with the following clarifications:

The valves discussed in items 1, 3 and 4 above were all included in the lnservice . Testing (IST) Program. The source of the violation results from inappropriate testing for the valves' safety function. CK-ES3339 and CK-ES3340 were being tested in the open direction, but not the closed direction. These valves have a safety function in both the open and closed direction. CV-1813, CV-1814, CV-1501, CV-1502 and CV-1503 were receiving the position indication test, but not a closed stroke time test, as required by Technical Specification 4.5.3.c, and Table 3.6.1.

BACKGROUND

During the. NRC Architectural/Engineering (A/E) Inspection, an NRC inspector identified the potential over pressure event of the High Pressure Safety Injection (HPSI) suction pipe associated with CK-ES3339 and CK-ES3340. Our subsequent IST Program reviews identified testing issues associated with CK-DMW400, CV-1813, CV-1814, CV-1501, CV-1502 and CV-1503 .

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Prompt operability determinations were performed to address the conditions. Preliminary cause evaluation and corrective actions addressed immediate program problems and implemented testing to verify compliance to applicable codes. Additionally, each deviation from IST and ISi requirements was reported under Licensee Event Report (LER) 97-013, "Violation of Technical Specification-6.5.7 for Failure to Fully Test Valves", dated December 9, 1997 and LER 97-013 Supplement 1, dated April 14, 1998.

REASON FOR VIOLATION

Minor weaknesses existed in the level of detail in ISi and IST bases and scoping documents which led to isolated errors in determining testing requirements. Although

. the ISi and IST programs exist to validate different functions, the scope of ISi is a key reference used by IST staff. A misalignment between design requirements and ISi boundaries contributed to this condition.

As a result, plant staff had an inadequate understanding of design details and therefore, did not consider all of the available information when determining the IST scope and testing requirements.

CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED

· 1. In each case,· upon discovery, an operability determination was made and the subject equipment was either declared inoperable until testing was performed, or was determined to be oper~ble per ASME Code or Technical Specifications requirements. All affected valveswhich were declared inoperable were subsequently tested and, ~ased on the results, declared operable within the 24 hour period allowed by Technical Specification 4.0.3. In retrospect, all valves remained capable of performing their safety functions.

-2. · Program elements were revised to incorporate appropriate functional and testing-.-----

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requirements.

Other ASME Section XI valves (check valves, MOV's arid AOV's) were reviewed for similar discrepancies in their Equipment Data Base designation and testing

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A TT AC HM ENT 1 REPLY TO NOTICE OF VIOLATION 50-255/98003-03



scope. Equipment Data Base revisions and procedure changes were initiated, as necessary. IST procedure changes and basis document changes were identified for enhancement of the IST program.

CORRECTIVE ACTIONS REMAINING TO AVOID FURTHER VIOLATIONS

1. The Palisades Equipment Database and IST basis documents for the affected valves will be revised.

2. A documented review of the ISi basis document will be completed to determine if similar conditions exist within the ISi program, as were identified in the IST program.

3 . The IST program and basis documents will be revised to enhance purpose, scope, and program descriptions. Revised testing requirements will be added for the subject valves. In addition, the IST program will be strengthened by adding additional technical reviews of surveillance tests.

DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED

Full compliance to the ISi and IST program requirements occurred on March 26, 1998 when the last set of valves were tested and declared operable.

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ATTACHMENT 2


DOCKET 50-255


FAILURE TO FOLLOW PROCEDURES AND UPDATE CALCULATIONS WHEN ANALYTICAL INPUTS CHANGED

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ATTACHMENT 2 REPLY TO NOTICE OF VIOLATION 50-255198003-04 FAILURE TO FOLLOW PROCEDURES AND UPDATE

CALCULATIONS WHEN ANALYTICAL INPUTS CHANGED



B. 1 O CFR Part 50, Appendix B, Criterion V, "Instruction, Procedures, and Drawings," requires, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and shall be accomplished in accordance with these procedures.

1. Administrative Procedure 9. 11, "Engineering Analysis," Revision 9, Section 6.1.5.c, states th~t an analysis shall be revised if analytical inputs changed .

Contrary to the above, as of November 14, 1997, the following analyses, which were activities affecting quality, were not revised when analytical inputs were changed or found to be in error:

a. The ambient air temperature of 94 degrees F assumed in Calculation EAFC-573-2, "Calculated Required Air Flow for Inverter/Charger Cabinet Cooling Fan," was not revised when deficiency report F-CG-91-072 indicated that a design basis temperature of 104 degrees F should have been used in the calculation.· (50-255/98003-04a(DRS))

b. An assumption regarding pipe break size was not updated in EA-A-NL-92-185-01, "Worst Case Operating Conditions for the LPCllSDC System

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MO Vs," to determine the effect of the motor operated valves to close against the break when a more conseNative pipe break assumption was used in a later analysis EA-C-PAL-95-1526-01, "Internal Flooding

_ _ Evaluation for Plant Areas Outside Containment," Revision 0. (50-. 255/98003-04b(DRS)) - - - - - - - .. - - - - ---- - -- ------ -·- -·- -- ---

Assumptions 5.9 and 5.10 of EA-A-NL-92-185-01, which stated that the High Pressure Safety Injection (HPSI) a.nd Low Pressure Safety Injection (LPSI) flows to the loops were approximately equal under post-accident

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ATTACHMENT 2 REPLY TO NOTICE OF VIOLATION 50-255/98003-04 FAILURE TO FOLLOW PROCEDURES AND UPDATE


conditions were not revised when flow values calculated in EA-SDW-95- . 001, "Generation of Minimum and Maximum HPSllLPSI System Performance Curves Using Pipe-Flo, " Revision 2, indicated that Assumptions 5.9 and 5.10 were incorrect. (50-255198003~04c(DRS))

The required LPSI injection check valve flows identified in EA-E-PAL-93-004E-01, "/ST Check Valve Minimum Flow rate Requirements to Support· Chapter 14 Events," Revision 0, were not revised after a new flow value was calculated in EA-SDW-95-001. (50-255/98003-04d(DRS))

Electrical calculation EA-ELECT-VOLT-13, "Palisades Loss of Coolant Accident With Offsite Power Available, " Revision 0, was not revised to

. reflect the load magnitudes identified in EA-ELECT-LDTAB-005, "Emergency Diesel Generator 1-1 & 1-2 Steady State Loading," Revision 4, and EA-SDW-95-001. (50-255/98003-04e(DRS))

This is a Severity Level IV Violation (Supplement 1).

CONSUMERS ENERGY RESPONSE

Consumers Energy Company agrees with this violation as written.

BACKGROUND

Recent industry experience associated with understanding and maintaining a plant's design basis have dictated the need for tighte;!r calculation control. The need to strengthen our controls was recognized during development of our response dated February 6, 1997 to the "Request for Information Pursuant to 50.54(f) Regarding Adequacy and Availability of Design Bases Information" letter. In addition, it was recognized that revising all affected calculations was unnecessary and not cost effective if the calculation was bounding.

The m~jorl!:t of r~J~y_~ot pli:;mt calculations a_nd analyse~ are reviewed and revised when an input is changed. However, every calculation which needed to be revised could not always be identified. While the need to locate and revise existing calculations:-ofrecord has been understood and incorporated into plant procedures, the programs and databases that identify calculations-of-record do not contain sufficient data to allow

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proper linking. When,Palisades was designed, the need to continually update calculations was not recognized. Therefore, administrative links back to the original calculation were seldom established and the methods for retrieval of inter-related calculations were difficult. Subsequent revisions to an implementing document or original calculation could be made without knowledge of this linkage. Although the current tools for identifying these links are not significantly different, current practices for calculation preparation and control are more rigorous. Recent changes which have strengthened Administrative Procedure (AP) 9.11, "Engineering Analysis", and related engineer training and qualification programs, have increased sensitivity of site engineers to the issue. These factors have helped compensate for the weaknesses in retrieval methods.

A Calculation Control Improvement Project was initiated to improve retrieval of calculations by establishing links between appropriate referenced calculations, and between calculations and implementing documents: This project will be complemented by the designation of an individual responsible for calculation~ and by providing specific engineering training focusing on the inter-relationships between plant changes.


The principle contributor to these errors was the lack of a strong method to identify the interdependent relationships between calculations and implementing documents. Engineers have needed to use other compem~atory practices to identify calculation linkages.

Another contributor was that the requirement in Administrative Procedure (AP) 9.11, "Engineering Analysis" to revise all affected calculations when their iriputs changed was unnecessary. For some of the examples cited, the preparer of the subsequent calculation was aware of the related calculation, but based on technical understanding of the issues, determined that the preceding calculation did not need to be revised. This was not in compliance with the established procedure.

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• CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED

1. Operability determinations were performed which verified that all equipment associated with the cited examples remained operable .

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AP 9.11 was revised to permit a bounding analysis to be generated without revising other issue-specific analyses. The revision also eliminated the requirement for input changes to other issue-specific analyses, if the changes are determined to be minor, and so documented in the subsequent analysis. Based on this revision, there is no need to revise calculations associated with the first four cited examples, as the calculations are determined to be bounded.


1. EA-ELEC-VOL T-13 will be revised to incorporate electrical load changes.

2. Continue implementation of the Calculation Control Improvement Project which includes identification of calculation inter-dependent relationships and sorting of calculations to distinguish calculations-of-record.


Full compliance for the first four cited examples was achieved on March 19, 1998 with revision of AP 9.11. Compliance for the final example, the failure to revise EA-ELECVOL T-13, will be achieved upon reyision of the calculation scheduled for September 15, 1999. Several design input calculations associated with this analysis need to be completed to support this analysis revision. This date is consistent with our 60 day response to the Architect/Engineering (NE) inspection report, dated March 2, 1998 .

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ATTACHMENT 3

CONSUMERSENERGYCOMPANY · PALISADES PLANT

DOCKET 50-255


FAILURE TO FOLLOW PROCEDURES AND REVIEW AND DOCUMENT THE ACCEPTABILITY OF SCAFFOLDING

INSTALLED IN THE VICINITY OF SAFETY RELATED EQUIPMENT

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ATTACHMENT 3 REPLY TO NOTICE OF VIOLATION 50-255/98003-05 FAILURE TO FOLLOW PROCEDURES AND REVIEW

AND DOCUMENT THE ACCEPTABILITY OF SCAFFOLDING INSTALLED IN THE VICINITY OF SAFETY RELATED EQUIPMENT


During an NRG inspection conducted from March 25, 1998, through April 10, 1998, six violations of NRG requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRG Enforcement Actions," NUREG-1600, violation 50-255198003-05is listed below:

2. Procedure MSM-M-43, "Scaffolding, " Revision 2, Section 5. 3. 1, "General, " requires that in addition to other requirements of this procedure, scaffolding constructed in the vicinity of safety related equipment shall conform to the following seismic requirements:

Suspension scaffolding ·shall not be used in any plant location which contains_ safety related equipment without prior engineering and approval and justification documented in Attachment 1, "data sheet, " Step 2. 6.

Responsible engineer shall provide justification and approval for any scaffold which deviates from the seismic requirements of this procedure, and shall document justification and approval in Attachment 1, "data sheet, " Step 2. 6.

Contrary to the above,· as of October 6, 1997, engineers had not reviewed the acceptability of scaffolding installed adjacent to the safety related safety injection and refueling water tank. In addition, on October 30, 1997, engineers had not reviewed the acceptability of scaffolding installed in the East engineering safeguards (ESG) room adjacent to safety related piping and documented it on Attachment 1, "data sheet," Step 2. 6.

This is a Severity Level IV Violation (Supplement 1). (50-255198003-05(DRS))

CONSUMERS ENERGY RESPONSE -

Consumers Energy Company agrees with this violation as written .

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AND DOCUMENT THE ACCEPTABILITY OF SCAFFOLDING INSTALLED IN THE VICINITY OF SAFETY RELATED EQUIPMENT

BACKGROUND

During the NRC Architect/Engineering (A/E) Inspection, an NRC inspector identified an improperly erected scaffold in the East Engineering Safeguards Room. The free standing scaffold was in close proximity to safety related equipment. The scaffold had been erected such that a six inch clearance was not maintained between the scaffold and the safety related equipment as required per Permanent Maintenance Procedure, MSM-M-43, "Scaffolding". This event was preceeded by asimilar event involving scaffolding next to the Safety Injection and Refueling Water (SIRW) Tank. An

, immediate action from this condition included a walk down of all scaffolding to ensure it was erected properly. The scaffold in the East Engineering Safeguards Room was inspected, but the inspection failed to identify the improper scaffold construCtion. The scaffold was removed upon identification.


Weaknesses existed in understanding and implementing procedural guidance provided in MSM-M-43 for the construction and inspection of scaffolds in the vicinity of safetyrelated and seismic equipment. Contributors associated with this condition include inadequate maintenance supervisor and repairman training on the requirements of MSM-M-43 and insufficient work order controls for scaffold construction and removal.


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MSM-M-43 expectation~ were reviewed with maintenance supervisors and repairmen.

After the second improper scaffold installation was found, walkdowns were conducted by Design Engineering to independently review implementation of scaffolding requirements, per MSM-M-43, in seismically sensitive areas. No additional non-compliances were identified ..

Maintenance planning enhancements have been made to require scaffold .. -- - construction-and removal .using work orders specific for.the maintenance-activity

associated with the scaffold.

Prior to and during the 1998 Refueling Outage, a Single Point of Control (SPOC) was assigned ownership of the scaffolding program. The.SPOC was responsible

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AND DOCUMENT THE ACCEPTABILITY OF SCAFFOl..DING INSTALLED IN THE VICINITY OF SAFETY RELATED EQUIPMENT

for scaffold scheduling, construction, inspection, tracking and removal. In addition, the SPOC provided an independent second level review of scaffold installations. This resulted in no additional non-compliances.


1. Enhanced MSM-M-43 scaffolding training will be incorporated into the Mechanical Maintenance Supervisors, Repairmen, and Contract Scaffolding Constructor Training Programs.


Consumers Energy has been in full compliance since November 1, 1997, when the scaffolding in question was removed. No further non-compliances were identified during independent walkdowns. ·

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ATTACHMENT 4


DOCKET 50-255


FAILURE TO FOLLOW PROCEDURES AND ADEQUATELY MAINTAIN THE REQUIRED SEPARATION DISTANCE BETWEEN AN

UNSECURED OPERATIONS STORAGE CABINET AND SAFETY RELATED PIPING AND VALVES IN THE WEST ESG ROOM

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3. Palisades Administrative Procedure 1.01, "Material Conditions _Standards and Housekeeping Responsibilities, "Revision 11, Appendix 2, requires that unrestrained and potentially damaging items which can topple should be separated from operable safety related equipment by a minimum horizontal distance equal to the height of the item plus five feet.

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Contrary to the above, on October 30, 1997, the inspectors observed an unsecured operations storage cabinet within 9 feet of safety related valves CV-0737 and CV-0747A in the West engineering safeguards rqom which was less than the required 11. 5 feet (6. 5 feet + 5 feet).

This is a Severity Level IV Violation (Supplement 1). (50-255198003-06(DRS))


Consumers Energy Company agrees with this violation as written.

BACKGROUND

During the NRC Architectural/Engineering (A/E) Inspection, an NRC inspector identified an unrestrained six and one-half foot high tool storage cabinet in the West Engineering Safeguards Room. Criteria from Administrative Procedure (AP) 1.01, "Material Conditions Standards and Housekeeping Responsibilities," Revision 11, Attachment 2, was used to determine the cabinet's aspect ratio, (height divided by the_base) .. Th.e ___ ·····- _ .. · .. : ___ _ cabinet's aspect ratio was found to be 3.9, which exceeded critical aspect ratio chart · value of 2.7, for the 570 foot elevation of the Auxiliary Building. AP 1.01, Attachment 2, Step 2 states, "unrestrained potentially damaging items, which can topple (aspect ratio value greater than chart value), should be separated from operable safety-related

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A TT A CHM ENT 4 REPLY TO NOTICE OF VIOLATION 50-255/98003-06



equipment by a minimum herizontal distance equal to the height of the item plus five feet." ·

Based on the aspect ratio of the cabinet, the potential existed for it to topple during a seismic event. Since the cabinet was six and one-half feet in height, it should have been a minimum of eleven and one-half feet from any safety-related equipment. However, the cabinet was found to be within nine feet of CV-0737 and CV-0747A, which are safety-related control valves. Therefore, the requirements of AP 1.01 were not met.


Although it was not determined who placed the cabinet near the safety-related equipment, it is most likely that it was placed by an Auxiliary Operator. Review of the Auxiliary Operator training program identified that AP 1.01 training is not performed . Consequently, the Auxiliary Operator could not be expeded to know about the requirements for unrestrained items near safety-related equipment.

Operations checklists ensure unrestrained items are properly controlled inside containment, however, they are not used to inspect for unrestrained items in areas outside containment.

AP 1.01 requirements state that materiel condition inspections should be performed on a routine basis and that the personnel conducting the inspections should consider the requirements for unrestrained items. These inspections are not always performed, nor

• are specific instructions available for implementing such inspections.


1. The subject cabinet was turned on its side in order to meet AP 1.01, Attachment 2 criteria, which eliminated this installation as potentially affecting safety-related equipment.--Subsequently, the cabinet was removed from the West Engineering Safeguards Room .

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UNSECURED OPERATIONS STORAGE CABINET AND SAFETY RELATED PIPING AND VALVES IN THE WESTESG ROOM ·

Additional walkdowns were performed to verify conformance to AP1 .01 requirements for unrestrained items. Deficiencies were documented and dispostioned.


1. The Auxiliary Operator training program will be revised to include training on unrestrained items and AP 1.01 requirements.

2. Plant start-up checklists for areas outside of Containment will be revised to inspect for unrestrained items.

3. Periodic walkdowns will be established to assess compliance.


Consumers Energy is now in full compliance .

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ATTACHMENT 5


DOCKET 50-255


FAILURE TO TEST OVERCURRENT RELAYS AS REQUIRED

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• ATTACHMENT 5 REPLY TO NOTICE OF VIOLATION 50-255/98003-09

FAILURE TO TEST OVERCURRENT RELAYS AS REQUIRED


During an NRG inspection conducted from March 25, 1998, through April 10, 1998, six violations of NRG requirements-were identified. In accordance with the "General Statement of Policy and Procedure for NRG Enforcement Actions," NUREG-1600, violation 50-255198003-09 is listed below: ·

4. Periodic and Predetermined Activity SPS025, "Bus 1 C Relay Testing," required testing of the overcurrent relays.

Contrary to the above, the licensee identified that as of October 27, 1997, the overcurrent relays for supply breakers 152-105 and 152-106 to Bus 1 C had not been overcurrent tested. On.June 28, 1995, during the 1995 refueling outage, work request 24416160 was issued for their calibration; however, no test results could be located to verify that testing had been completed.

This is a Severity Level IV Violation (Supplement 1 ). (50-255/98003-09(DRS))

• CONSUMERS ENERGY RESPONSE

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Consumers Energy Company agrees with this violation _as wr.itten.

BACKGROUND

As part of preventive maintenance program, testing is performed on Bus 1 C relays every three years or every other refueling outage. The test frequency was based on · engineering judgement and was not code, standard or license based. The preventive maintenance test for the Bus 1 C relays was .scheduled to be completed during the 1995 refueling outage. ·

Relay testing associated with SPS025 was performed during the 1995 refueling outage, however, the Bus 1 C breakers 152-105 ~md 152-106 overcurrent relay calibration was missed. The Work Order closeout review failed to recognize the subject relays had not been tested.


The main reason for the violation was determined to be the failure of individuals performing SPS025 to recognize the overcurrent relays were not calibrated during the performance of the work activity. Weaknesses in specifically identifying the calibration

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. ATTACHMENT 5 REPLY TO NOTICE OF VIOLATION 50-255/98003-09

FAILURE TO TEST OVERCURRENT RELA VS AS REQUIRED

and verification requirements while developing the work order plan for perlorming SPS025 contributed to this condition. Subsequent work order completion reviews were inadequate in that they did not identify this condition. As a result, the work order was closed without the relays tested.


1. The relays were declared operable based upon satisfactory past operation and performance data.


1. Maintenance personnel, planners and supervisors will be briefed on the expectations for preparing and implementing work orders associated with protective relay testing.

2. Test the overcurrent relays for 152-105 and 152-106. This action is scheduled for completion by December 31, 1998 .


Consumers Energy is in full compliance as of October 28, 1997 when an operability determination was performed and it was determined that the relays were operable

· based upon satisfactory past operation and performance data.

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ATTACHMENT 6


DOCKET50-255.

REPLY TO NOTICE OF VIOLATION 50-255/98003-07a AND 7b

FAILURE TO CORRECTLY CONSTRUCT A PORTION OF THE CONTAINMENT SUMP IN ACCORDANCE WITH THE DESIGN DRAWINGS

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FAILURE TO CORRECTLY INSTALL INSTRUMENT TUBING FOR THE HPSI AND LPSI FLOW TRANSMITTERS WITH THE CORRECT SLOPE

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ATTACHMENT 6 ~EPL Y TO NOTICE OF VIOLATION 50-255198003-07a AND 7b

FAILURE TO CORRECTLY CONSTRUCT A PORTION OF THE CONTAINMENT SUMP IN ACCORDANCE WITH THE DESIGN DRAWINGS AND



During an NRC inspection conducted from March 25, 1998, through April 10, 1998, six violations of NRG requirements were identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violation is listed below:

C. 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50. 2 and as specified in the license application, for those, structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions .

Drawing M-7 4, "Underground Piping Reactor Building, " Sheet 1, Revision 10, and Drawing C-155, "Reactor Building Refueling Cavity and Sump Liner," Sheet 2, Revision 12, specify screens on the inlets of two vent pipes connecting the 590-ft elevation ·of the containment and the containment sump.

UFSAR Section 6.4.2.3 states that the design of the spray nozzles was reviewed to confirm that the spray nozzles are not subject to clogging from debris entering the recirculation system through the containment sump screens.

Drawings J-F-152, "Flow Instrument Above Line WNents - Liquids," Revision 1 and J-F-153, "Flow Instrument Above Line WNents - Liquids," Revision O, specify a tubing slope of one inch per foot of instrument tubing run for flow transmitters in the High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection (LPSI) systems.

Contrary to the above:

a. As of September 26, 1997, two vent pipes which connected the containment sump to the 590-ft elevation of the containment did not have screens installed. This piping configuration resulted in a pathway for . debris to enter the recirculation system without being filtered by the

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ATTACHMENT 6 REPLY TO NOTICE OF VIOLATION 50-255/98003-07a AND 7b



containment sump screens and had the potential to clog the containment spray nozzles. (50-255198003-07a(DRS))

b. As of October 24, 1997, instrument tubing to HPSI flow transmitters FT-0308, FT-0310, FT-0312, and FT-0313 and LPSI flow transmitters FT-0307, FT.,.0309, FT-0311, and FT-0314 was notinstal/ed as specified in

·Drawings J-F-152 and 153, in that it did not have the specified one-inch per foot slope. (50-255198003-07b(DRS))

This is a Severity Level IV Violation (Supplement 1 ).


Consumers Energy Company agrees with this violation 50-255/98003-07a, as written, with one exception. As a correction to the NOV description, design drawings M-74, "Underground Piping Reactor Building", Sheet 1, Revision 10, and Drawing C-155, "Reactor Building Refueling Cavity and Sump Liner,'' Sheet 2, Revision 12 did not identify, nor have they in past revisions identified, the two containment sump vent pipes as having screens pn the pipe inlets.

Consumers Energy contests violation 50-255/98003-07b, as described in Section II of this reply.

I. BACKGROUND FOR 50-255/98003-7a

During a review of the containment sump vent lines by station Design Engineering personnel, it was recognized that the containment sump vent piping enters the containment sump behind the main sump suction piping screens. The other end of the vents open into containment at elevation 595 feet. The currently predicted

. maximum containment flood level is approximately 597 feet or two feet above the vent outlets. With the vent outlets covered with water following a Design Basis

_ Accident,. an .. unscreened flow path exists between the containment-and-the-24--inch diameter suction lines for the Engineering Safeguards Pumps. The vent was considered a valid flow path for debris larger than the maximum allowable size to enter the containment sump and subsequently, the Emergency Core Cooling and Containment Spray Systems. This condition affects both suction lines from the

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containment sump. The purpose of the vent lines is to vent the sump during. initial sump fill and appears to allow relieving of the differential pressure across containment sump screens, if they are blocked by debris.


A good understanding of the containment sump design basis exists, however, some of the design detail for the containment sump attributes has not been documented. Engineered Safeguards System Design Basis Documents historically have not included the containment sump design attributes.

Containment sump vent lines have not had protective screens since original plant construction, however, opportunities have existed for earlier recognition of this as being a potential problem. Several revisions to the containment flood calculation have been performed to capture design conservatism and design changes. The results of those revisions increased the flood level above the vents, however, the impact on sump vent lines was not recognized.


1. An immediate operability recommendation analyzed several containment attributes that could potentially exist, and affect, containment sump equipment capabilities during the most conservative containment flooding scenario. Reasonable assurance was gained to conclude that the containment sumi;:>, Emergency Core Cooling System, and Containment Spray System were operable.

2. A Temporary Modification was implemented t.o cover the containment sump vents with a mesh screen. The Temporary Modification was converted to a permanent installation, using an approved design change during the 1998 refueling outage in May 1998.

3. · -- - The station implemented an Independent Safety and Design Review (S&DR) Group in early 1996 to address similar past design control issues. This group independently reviews design documents for conformance to design basis. The record revision of the containment flood analysis was performed before this group was commissioned. S&DR has proven to be effective in fulfilling its objective .

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FAILURE TO CORRECTLY INSTALL INSTRUMENT TUBING FOR THE HPSI AND LPSI FLOW TRANSMIITERS WITH THE CORRECT SLOPE


1. The Containment Spray Design Basis Document DBD-2.03 will be revised to add containment sump information. Items such as containment sump design basis, function, margins, etc., are to be included. This action will address issues vital to the function of the ·Engineered Safety Features following a LOCA.

2. Commitments have been identified in our 1 OCFR50.54(f) response, dated February 6, 1997, that will further ensure adequacy and availability of design basis information.


Full compliance was obtained upon the installation of screens on the containment sump vent lines on October 27, 1997.

11. BASIS FOR DISPUTING 50-255/98003-7b

During the Architect/Engineering (AE) inspection in the Fall of 1997, NRC personnel performing a detailed review of the High Pressure Safety Injection (HPSI) and Low Pressure Safety Injection (LPSI) systems, identified that the flow transmitters for the HPSI and LPSI pumps do not have adequate sensing line slope. This was· based on review of generic design document drawings J-F-152 and J-F-153 and photographs of the sensing lines. Physical verification of the sensing line configuration could not be performed at that time, since the installations in question are located in an area that is inaccessible while the station is operating. As a conservative measure, station personnel evaluated the acceptability of the performance of the HPSI and LPSI flow indications based on worst case installation configuration. Subsequent information, obtained by performance of walkdowns during the recent refueling outage, confirmed that the

-HPSi-<:ind LPSI ·flow transmitter sensing lines are appropriately sloped. This information was not available during the A/E inspection. Consumers Energy is now reporting the results and basing our dispute of the A/E inspection finding on them .

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Detailed walkdowns were performed of various safety significant instrument installations to verify that slope requirements were being met. Additionally, maintenance history was researched for .several safety significant applications to identify repetitive problems related to improper sensing line slopes. No · deficiencies were· noted.

As a clarification to the NOV description, design drawings J-F-152 and J-F-153 are generic transmitter installation details for sensing lines and their slope requirements. For these original HPSI and LPSI transmitter installations, design details do not exist. In general, the HPSI and LPSI sensing lines meet the generic drawing design requirements of one-inch per foot slope. Design details are being updated to reflect the installed configuration.

There is no safety impact or adverse consequences associated with the existing HPSI and LPSI transmitter flow sensing line installation, as the sensing lines are appropriately constructed. In addition, a review of past performance has shown reliable operation in over 25 years of service .

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ATTACHMENT 7 .


DOCKET 50-255

REPLY TO NOTICE OF DEVIATION 50-255/98003-02

DEVIATION FROM A RG 1.97 COMMITMENT

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ATTACHMENT 7 REPLY TO NOTICE OF DEVIATION 50-255/98003-02


NRC NOTICE OF DEVIATION

During an NRG Inspection conducted on March 25 through April 10, 1998, two deviations from commitments were identified. In accordance with the "General Statement of Policy and Procedure for NRG Enforcement Actions," NUREG-1600, deviation 50-255198003-02 is listed below .

. A. NRG letter to Consumers Power Company dated July 19, 1988, entitled "Palisades Plant - Response to Generic Letter 82-33 Conformances to Regulatory Guide 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following An Accident" allowed the use of temperature instruments to monitor CCW flow.

Regulatory Guide (RG) 1. 97 stated a range for CCW flow instrumentation of 0-110 percent of flow.

FSAR Appendix 7C, "Regulatory Guide 1.97 Instrumentation," page 21of29, which described the commitment to RG 1.97, Revision 3, stated the range of the temperature instruments used to measure CCWflow (TE-0912 and TE-0913) was 0-180 degrees F.

Contrary to the above, CCW flow could not be measured from 0-11 O percent of flow using the listed temperature instruments because their indication range was 0-180 degrees F and recent sensitivity studies indicated that the outlet temperature of ccw from the shutdown cooling heat exchanger would be 184 degrees F. (50-255/98003-02(DRS)) ·


Consumers Energy Company agrees with this deviation as written.

BACKGROUND

The original-maximum design temperature for the Component Cooling-Water (COW) -------· --- -- ----system was 140°F. Temperature Elements TE-0912 and TE-0913 are original Plant installed instrumentation that were designed to encompass the original maximum CCW temperature. In 1988, their indicated range (0 to 180°F) was used to comply with the Regulatory Guide 1.97 requirement for confirming CCW flow (0 to 110 percent). For the

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11 O percent criteria, the 0 to 180°F range would be acceptable for an actual maximum CCW temperature of approximately 163°F.

During preparation for the ArchitecVEngineering (AE) Inspection, it was discovered that Loss of Coolant Accident (LOCA) temperatures under the latest containment analysis were greater than the CCW System design temperature. Design Basis Document (DBD) 1.01 states that th.e maximum CCW System design temperature is 140°F, but analysis of the present LOCA conditions predicted a maximum of 184°F. Several previous analyses of LOCA conditions had been performed, but none of them had looked at the effect of the maximum temperature on the RG 1.97 stated range of 0-110 percent flow, or the 0-180°F range of the temperature instruments (Tl-09.12 and Tl-0913) used to confirm flow.

REASON FOR DEVIATION

The cause for the deviation was the failure to recognize the relationship between analysis inputs/outputs and design basis values for the systems associated with the analysis. Therefore, the DBD owner, system engineer, and EOP owner did not have an opportunity to evaluate the impact of the analysis on their systems.


1. A new Engineering Analysis was performed to determine the most limiting conditions for the CCW system and the maximum CCW temperature that would be seen during a LOCA. The results showed the maximum CCW temperature to be 180 ° F, which confirmed that an issue still existed for-meeting the maximum 163°F value.

CORRECTIVE ACTIONS REMAINING TO AVOID FURTHER DEVIATIONS

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The range of Tl-0912 and Tl-0913 will be changed to encompass the results of the new· Engineering Analysis.·

Continue implementation of the Calculation Control Improvement Project which - -includes-training for-technical support staff on subtle errors made involving

calc.ulation updating .

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Full compliance will be achieved after the modification to the range of Tl-0912 and Tl-0913, currently scheduled to be completed by the end of 1998 .

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ATTACHMENT 8


DOCKET 50-255

REPLY TO NOTICE OF DEVIATION 50-255/98003-08


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NRC NOTICE OF DEVIATION

During an NRG Inspection conducted on March 25 through April 10, 1998, two deviations from commitments were identified. In accordance with the "General Statement of Policy and Procedure for NRG Enforcement Actions," NUREG-1600, deviation 50-255198003-08 is listed below.

B. Licensee Jetter to the NRG dated January 24, 1978, stated that the recommendations of Regulatory Guide 1. 6 would be implemented, in that, no provision would exist for automatically transferring loads between redundant power sources.

NRG Safety Evaluation Report dated April 7, 1978, confirmed this commitment.

Facility Change FC-364, "Feeder Change for Instrument Bus Y-01, Revision 0, implemented this commitment and powered bus Y-01 from Motor Control Center (MCC) 1 and non-safety related MCC 3.

Contrary to the above, FC-854 moved the backup power source from MCC 3 to MCC 2, a redundant power source, which resulted in Bus Y-01 being able to · automatically transfer between two safety related busses via transfer switch Y-50. (50-255198003-0B(DR S))

CONSUMERS ENERGY COMPANY RESPONSE

Consumers Energy contests this Notice of Deviation.

BASIS FOR DISPUTING THE NOTICE OF DEVIATION

NUREG-1600, "General Statement of Policy and Procedure for NRC Enforcement Actions," Section Vl.D, "Related Administrative Actions," sfates that the NRC expects licensees to adhere to any obligations and commitments resulting from formal .enforcement actions, or from administrative actions on the part of th.e NRC such as Demands for Information. Consumers Energy Company agrees with this expectation and concludes thatthis expectation was met in this case. We have maintained our commitment to assure that redundant safety related power sources cannot both be affected by a fault on the instrument bus. ·

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•• ATTACHMENT 8 REPLY TO NOTICE OF DEVIATION 50-255/98003-08


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In our letter of January 24, 1978, Consumers Energy stated that a proposed modification would eliminate the concern for the probability of a fault on the instrument bus (Y-01) being transferred from one safety group bus to the other, by the action of the transfer switch (Y-50), with a possible consequence of the loss of both safety group buses. Consumers went on to state that the recommendations of Regulatory Guide 1.6 would be met in regard to the instrument system by the proposed modification.

As of the January 24, 1978 letter, the instrument bus was supplied by either of two safety related power sources; motor control center (MCC) -1, and MCC-2; with the supply normally from MCC-1 and an automatic transfer to backup source MCC-2 performed by Y-50 upon loss of MCC-1 voltage. lr:i the January 24, 1978 letter, Consumers stated that the MCC-2 feed would be replaced by ,a feed from non-safety related source MCC-3. /

By letter dated April 7, 1978, the NRC acknowledged our proposed modification indicating that it would eliminate the possibility of transferring a fault to a redundant MCC. The NRC also stated that the proposed modification would satisfy the provisions of Regulatory Guide 1.6, and quoted Regulatory Guide 1.6 stating - "No provision should exist for automatically transferring loads between redundant power sources."

. Facility Change (FC) -364 was subsequently implemented at Palisades to replace safety related feed MCC-2 to the instrument bus with non-safety feed MCC-3 to assure that a fault on the instrument bus would not affect both safety related electrical divisions.

In 1990, Palisades implemented FC-854, to return the backup power source from MCC-3 to MCC-2 to regain power supply reliability for instrument bus loads. In addition, FC-854 installed an isolation fuse in each MCC feed (both MCC-1 and MCC-2 feeds) in series with the feed's protective breaker. This feature provides diverse and redundant protection for each MCC; serving to assure that the original concern remained satisfied in that a transfer of an instrument bus fault to a redundant safety related MCC could not occur.

The 1 O CFR 50.59 safety evaluation for FC-854 stated that the addition of an isolation fuse in each MCC feed would prevent a failure of instrument bus Y-01 equipment from affecting both MGC-1 and-MCC-2, and also stated that FSAR redundancy-requirements-- -- ____ _ were satisfied. As a result, the safety evaluation concluded that an unreviewed safety question did not exist and prior NRC approval was not required. A description of the implemented modification was transmitted to the NRC in our Annual Report of Facility Changes, Tests and Experiments dated April 2, 1991 .

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We have maintained our commitment to assure that redundant safety related power sources cannot both be affected by a fault on the instrument bus. We did, however, change, through a codified process, our more specific commitment contained in our January 24, 1978 letter which described the manner in which we would protect both redundant power sources. Since the installation of FC-854, we utilize diverse and redundant isolation in each safety related power supply feed to the instrument bus rather than the assignment of a non-safety related feed as the backup power source which had been implemented earlier by FC-364.

In 1990, at the time.that FC-854 was implemented and qur more specific commitment was changed, neither industry guidance nor regulatory requirements existed which defined how changes in commitments should be processed. As a result, the changes made by FC-854 to reassign the backup power source to MCC-2 and provide an isolating fuse in each feed to the instrument bus, were evaluated in accordance with 10 CFR 50.59 and reported to the NRC in accordance with 10 CFR 50.59 and 10 CFR 50.71 (e).

By the end of 1995,_ the Nuclear Energy Institute (NEI) had developed specific commitment management guidance in the form of a guideline entitled "Guideline for Managing NRC Commitments" with Revision 2 of the guideline issued December 19, 1995. In SECY 95-300, dated December 20, 1995, the NRC stated that the NEI guidelines were an acceptable guide for licensees to follow for managing and changing their commitments to the NRC. By letter to NEI, dated January 24, 1996, the NRC endorsed the use of NEl's "Guideline for Managing NRC Commitments", stating that the

· guideline provides a logical method for evaluating commitments for possible modification.

To evaluate retrospectively the acceptability of our change in specific commitment made through FC-854, the guidance provided in the aforementioned NEI guideline is used. Guideline Step 1 entitled "Is There a Codified Change Process for the Commitment?" states that if a complete 10 CFR 50.59 review determines that an unreviewed safety question does not exist, licensees may make the change and provide a description of the change to the NRC annually or coincident with filing FSAR updates. The FC-854 change in specific commitment was made in accordance with today;s guidelines and represents appropriate commitment change management.

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