LR-N980316, Responds to NRC Re Violations Noted in Insp Rept 50-354/98-05.Corrective Actions:Interaction Between Nuclear Fuels,Sys Engineering & Outage Mgt Has Been Proceduralized & Decay Heat Removal Method Evaluated IAW 10CFR50.59
| ML20236M586 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/06/1998 |
| From: | Keiser H Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-354-98-05, 50-354-98-5, LR-N980316, NUDOCS 9807140147 | |
| Download: ML20236M586 (8) | |
Text
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Public Service Electric and Gas a Company Hir:Id W. Keiser Pubhc Service Electnc and Gas Company PO Box 236, Hancocks Bndge. NJ 08038 609-339-1100 Cruef Nuclear Officer & Presg3ent Nucleat Business Urut LR-N980316 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REPLY TO NOTICE OF VIOLATION INSPECTION REPORT 354/98-05 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 Pursuant to the provisions of 10 CFR 2.201, Public Service Electric and Gas Company (PSE&G) hereby submits a reply to the Notice of Violation (NOV) issued to the Hope Creek Generating Station in a letter dated June 4,1998. l The PSE&G response for this violation is contained in the Attachment to this letter. If you have any questions or comments on this transmittal, please contact Paul Duke at (609)339-1466.
Sincerely, i M Attachment
/y !l l 9807140147 980706 PDR ADOCK 05000354 G PDR
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l JUL 0 61998 Document Control Desk ! LR-N950316 l
C Mr. H. Miller, Administrator - Region l l- U. S. Nuclear Regulatory Commission j
- 475 Allendale Road
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l King of Prussia, PA 19406 i
! I Mr. R. Ennis, Licensing Project Manager- Hope Creek U. S. Nuclear Regulatory Commission l One White Flint North 11555 Rockville Pike !
Mail Stop 14E21 Rockville, MD 20852 l
Mr. S. Pindale (X24) i USNRC Senior Resident inspector- HC i Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P. O. Box 415 Trenton, NJ 08625 I .
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l l ATTACHMENT RESPbNSE TO NOTICE OF VIOLATION INSPECTION REPORT NO. 50-354/98-00 HOPE CREEK GENERATING STATION I
DOCKET NO. 50-354 A.10 CFR 50.59 Violation
- 1. DescripUon of the Notice of Violation 10 CFR 50.59, " Changes, tests and experiments," in part, permits the licensee to make changes to its facility and procedures as described in the final safety analysis report (FSAR) and conduct tests or experiments not described in the safety analysis report without prior Commission approval provided the change does not involve a change in the technical specifications or an Unreviewed Safety Question (USQ). The licensee shall maintain records of changes in the facility and these records must include a written safety evaluation which provides the bases for the determination that the change does not involve a USO.
FSAR Section 9.1.3.2.3 establishes that the design and operation of the fuel pool cooling and cleanup systems for the decay heat associated with a full core offload is based, in part, on the operation or availability of the residual heat removal (RHR) system to augment the fuel pool cooling and cleanup (FPCC) system.
Contrary to the above, during refueling outage RF03 in December 1990, the licensee did not maintain the RHR system in operation or available to augment the FPCC system which represented a change to the facility as described in the FSAR and did not perform a review of this change to demonstrate that the change did not involve a USQ.
This is a Severity Level IV violation (Supplement 1).
- 2. Reply to Notice of Violation PSE&G agrees with the violation.
- 3. Reason for the Violation PSE&G attributed the cause for this violation to inadequate procedures for outage reviews and controls. Alternate means of decay heat removal were evaluated before the RFO3 full core offload to ensure sufficient decay heat l removal capacity. However, the requirement to compare the alternate decay heat removal metbads with those described in the Hope Creek Updated Final Safety Analysis Report (UFSAR) and to evaluate deviations in accordance with 10 CFR 50.59 was not recognized.
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Atta3hmont LR-N980316
- 4. Corrective Steps that Have Been Taken and Results Achieved The interaction among Nuclear Fuels, System Engineering and Outage Management has been proceduralized. The station outage risk management procedure was revised to include guidance on the development of decay heat load estimates and heat-up curves for outage planning. The guidance includes verification of adequate decay heat removal capability throughout the outage schedule.
For the full core offload performed during refueling outage RFO7, completed in December 1997, the decay heat removal method was evaluated in accordance with 10 CFR 50.59 and found not to involve an Unreviewed Safety Question.
- 5. Corrective Steps that Wdi Be Taken to Avoid Further Violations No additional corrective actions are planned.
- 6. Date When Full Compliance Will be Achieved Hope Creek achieved full compliance when the RHR system was returned to available status and the core reload was completed during RFO3.
B.10 CFR 50 Appendix B Criterion XI Violation
- 1. Description of the !Jotice of Violation 10 CFR 50 Appendix B Criterion XI requires, in part, that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service be identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The test program shall include, as appropriate, proof tests prior to installation and operational tests during nuclear power plant operation of structures, systems, and components.
Contrary to the above, two examples of inadequate testing requirements associated with a design change modification to the Hope Creek safety-related control area chilled water syste n chillers were identified as follows:
(1) As of April 7,1998, a complete proof test prior to installation and an i operational test had not been performed to verify that check valves 1KBV-1243 through 1KVB-1250 [ sic) would provide a relatively leak tight boundary and ensure that the backup safety-related pneumatic supplies for the chiller condenser cooling water pressure control valves would remain available for four hours after a loss of power event.
(2) On April 8,1998, the backup safety-related pneumatic pressure regulators (1KBPCV-11464, -11466, and -11467) for the chiller condenser cooling water Page 2 of 6
Attathm:nt LR-N980316 pressure control valves were found set below minimum design requirements.
Operational tests had also not been performed to ensure that pressure regulators 1KBPCV-1164 [ sic) through 1KBPV-1171 [ sic] would remain properly set in accordance with design requirements.
This is a Severity Level IV violation (Supplement 1).
- 2. Reply to Notice of Violation PSE&G agrees with the violation.
- 3. Reason for the Violation Example (1)
PSE&G attributed the cause of the failure to establish inservice testing (IST) requirements and to periodically perform ISTs to personnel error. Personnel preparing the design change package (DCP) and performing design specialty reviews did not ensure the DCP was reviewed by the IST group. Deficiencies in the standard DCP format contributed to the violation. The design interface record had a single signoff for IST and for valve programs.
As part of the post-modification testing, an external leak check was performed and the valves were functionally tested. Both tests were satisfactory. However, since the design change packages that added the check valves for the backup pneumatic supply were not reviewed by the IST group, IST requirements for the valves were not established.
Example (2)
PSE&G attributed the cause of the failure to maintain the backup pneumatic pressure regulator settings to personnel error, most likely after the modification i was completed. The regulator settings were verified as part of the post-modification testing. The most likely scenario is misadjustment by an operator during rounds or during a routine maintenance activity. Deficiencies in the ,
procedure changes and in operator training for the design change package i contributed to the violation.
- 4. Corrective Steps that Have Been Taken and Results Achieved Example (1)
- a. The check valves were tested satisfactorily.
- b. IST procedures for the check valves have been developed.
- c. The format for the design change interface record was revised to require a separate signoff for the IST review.
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Atta:hm:nt LR-N980316 l
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' . A detailed roll out of this event and its implications was provided to affected design engineering personnel.
- e. Personnelinvolved were counseled concerning performance in this event.
Example (2)
- a. The backup pneumatic supply regulator settings were restored to their required values.
- b. Interim guidance on the operation of the backup pneumatic supply system was provided to operators, i
- c. IST procedures for the check valves have been developed. The procedures include periodic verification of the pressure regulator settings for the backup pneumatic supply.
- 5. Corrective Steps that Will Be Taken to Avoid Further Violations Example (1)
No additional corrective actions are planned.
Example (2)
- a. This violation will be incorporated in operator continuing training by September 1,1998.
- b. Lessons learned from this violation will be communicated to Engineering personnel by September 30,1998. ;
- 6. Date When Full Compliance Will be Achieved Example (1)
Hope Creek achieved full compliance on April 8,1998 when the inservice testing l
was performed satisfactorily on the backup pneumatic supply check valves. The valves have been added to the IST program.
l Example (2) ,
l Hope Creek achieved full compliance on April 8,1998 when the backup pneumatic supply regulator settings were restored to their required values. i j
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Attanhm:nt LR-N980316 -
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C.10 CFR 50 Appendix B Criterion XVI Violation l
- 1. Description of the Notice of Violation I
10 CFR Apnendix B Criterion XVI (Corrective Action) requires, in part, that {
measures shall be established to assure that conditions adverse to quality, such 1 as failures, malfunctions, deficiencies, deviations, and nonconformances are promptly identified and corrected.
i' Contrary to the above, on December 10,1997, PSE&G engineers determined l L that the minimum cooling water inlet temperature for the safety-related control I area chilled water system chillers should be changr J in a more limiting direction to 70 degrees Fahrenheit from 55 degrees Fahrenneit. On April 9,1998, the l operations department management, still unaware of any necessary change to the minimum allowed cooling water temperature, used 55 degrees Fahrenheit as a basis for determining inoperability when they made a four-hour event
! notification to the NRC. Hope Creek abnormal operating procedure, Loss of Instrument Air and/or Service Air, HC.OP-AB.ZZ-0131(Q) - Rev.14, and pending :
change, HFSAR 97-080, to the Hope Creek Updated Final Safety Analysis b
. Report (UFSAR) also incorrectly stated that 55 degrees Fahrenheit was the minimum cooling water temperature below which the safety-related backup pneumatic supply needed to remain operable. The change in minimum cooling water inlet temperature to a more limiting value was not corrected until May 7, 1998, when guidance was provided to operators specifying the new 70 degrees Fahrenheit minimum cooling water temperature.
This is a Severity Level IV violation (Supplement 1).
- 2. Reply to Notice of Violation PSE&G agrees with the violation.
- 3. Reason for the Violation I
PSE&G attributed the cause for this violation to personnel error, in December, 1997, the responsible engineer concluded that the minimurn Safety Auxiliaries Cooling system (SACS) temperature for Control Room chiller operation with full J
. SACS flow is higher (more limiting) than the minimum temperature used for design of the backup pneumatic supply modification. The 55 degrees Fahrenheit temperature was an appropriate limit for a fully loaded chiller; but it is more conservative to assume that the chiller is lightly loaded. The responsible l engineer, who is no longer employed by PSE&G, recognized the need for l corrective action but did not initiate an Action Request as required by PSE&G's Corrective Action Program to ensure the non-conservative design assumption was reviewed for its effect on chiller operability.
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Attachmont LR-N980316
- 4. Corrective Steps that Have Been Taken and Results Achieved
'a. An Action Request to document this condition was initiated.
- b. The backup pneumatic supply was restored and the Control Room chillers were returned to OPERABLE status on April 8,1998.
- c. A detailed roll out of this event and its implications was provided to affected design engineering personnel.
- 5. Corrective Steps that Will Be Taken to Avoid Further Violations
- a. An evaluation to determine the correct minimum SACS temperature for chiller operation without the Instrument Air system or backup pneumatic supply will be completed by August 21,1998. Temporary administrative controls are in place to ensure the backup pneumatic supply remains in service when SACS temperature is less than 70 degrees Fahrenheit (for Control Room chillers) or 62 degrees Fahrenheit (for 1E Panel Room chillers).
- b. Operating procedures will be revised as necessary by September 18,1993 to t
include the results of the evaluation described above.
- c. Lessons learned from this violation will be communicated to Engineering personnel by September 30,1998.
- 6. Date When Full Compliance Will be Achieved Hope Creek achieved full compliance on April 8,1998 when the backup pneumatic supply to the chiller pressure control valves was restored.
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Page 6 of 6 w____________________.
. . -; / . Public Service Electnc and Gas
- Company Hirdd W. Keiser Pubhc Service Electne and Gas Company PO Box 236, Hancocks Bndge, NJ 08038 609-339-1100 Chief Nuclear Ottecer & President Nuclear Business Unit LR-N980316 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REPLY TO NOTICE OF VIOLATION INSPECTION REPORT 354/98-05 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 Pursuant to the provisions of 10 CFR 2.201, Public Service Electric and Gas Company (PSE&G) hereby submits a reply to the Notice of Violation (NOV) issued to the Hope Creek Generating Station in a letter dated June 4,1998.
4 The PSE&G response for this violation is contained in the Attachment to this letter. If :*
you have any questions or comments on this transmittal, please contact Paul Duke at (609) 339-1466. ,
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Sincerely, I
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W JUL 061998 Document Control Desk l LR-N980316 ~.w i
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C Mr. H. Miller, Administrator- Region i U. S. Nuclear Regulatory Commission 475 Allendale Road l
King of Prussia, PA 19406 Mr. R. Ennis, Licensing Project Manager - Hope Creek U. S. Nuclear Regulatory Commission One White Flint North 1
11555 Rockville Pike Mail Stop 14E21 Rockville, MD 20852 Mr. S. Pindale (X24)
USNRC Senior Resident inspector - HC Mr. K. Tosch, Manager IV I Bureau of Nuclear Engineering P. O. Box 415 Trenton, NJ 08625 l
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. . -c .<- ATTACHMENT RESPONSE TO NOTICE OF VIOLATION INSPECTION REPORT NO. 50-354/98-05 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 l
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! A.10 CFR 50.59 Violation i
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- 1. Description of the Notice of Violation 10 CFR 50.59, " Changes, tests and experiments," in part, permits the licensee to make changes to its facility and procedures as described in the final safety analysis report (FSAR) and conduct tests or experiments not described in the safety analysis report without prior Commission approval provided the change does not involve a change in the technical specifications or an Unreviewed Safety Question (USQ). The licensee shall maintain records of changes in the facility and these records must include a written safety evaluation which provides the bases for the determination that the change does not involve a USQ.
FSAR Section 9.1.3.2.3 establishes that the design and operation of the fuel pool cooling and cleanup systems for the decay heat associated with a full core
~
offload is based, in part, on the operation or availability of the residual heat removal (RHR) system to augment the fuel pool cooling and cleanup (FPCC) ,
l system.
Contrary to the above, duripg refueling outage RF03 in December 1990, the licensee did not maintain the RHR system in operation or available to augment the FPCC system which represented a change to the facility as described in the FSAR and did not perform a review of this change to demonstrate that the change did not involve a USQ.
This is a Severity Level IV violation (Supplement 1).
- 2. Reply to Notice of Violati _
PSE&G agrees with the violation.
- 3. Reason for the Violation l l
PSE&G attributed the cause for this violation to inadequate procedures for j l outage reviews and controls. Alternate means of decay heat removal were j evaluated before the RFO3 full core offload to ensure sufficient decay heat l
removal capacity. However, the requirement to compare the attemate oecay l heat removal methods with those described in the Hope Creek Updated Final Safety Analysis Report (UFSAR) and to evaluate deviations in accordance with l 10 CFR 50.59 was not recognized. !
Page 1 of 6 1
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Attashment LR-N980316
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- 4. Corrective Steps that Have Been Taken and Results Achieved The interaction among Nuclear Fuels, System Engineering and Outage Management has been proceduralized. The station outage risk management procedure was revised to include guidance on the development of decay heat load estimates and heat-up curves for outage planning. The guidance includes verification of adequate decay heat removal capability throughout the outage schedule.
For the full core offload performed during refueling outage RFO7, completed in December 1997, the decay heat removal methoo was evaluated in accordance with 10 CFR 50.59 and found not to involve an Unreviewed Safety Question.
- 5. Corrective Steps that Will Be Taken to Avoid Further Violations No additional corrective actions are planned.
- 6. Date When Full Compliance Will be Achieved Hope Creek achieved full compliance when the RHR system was returned to available status and the core reload was completed during RFO3.
B.10 CFR 50 Appendix B Criterion XI Violation
- 1. Description of the Notice of Violation 10 CFR 50 Appendix B Criterion XI requires, in part, that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service be identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The test program shall include, as appropriate, proof tests prior to installation and operational tests during nuclear power plant operation of structures, systems, and components.
l Contrary to the above, two examples of inadequate testing requirements ass-ociated with a design change modification to the Hope Creek safety-related control area chilled water system chillers were identified as follows:
(1) As of April 7.1998, a complete proof test prior to installation and an operational test had not been performed to verify that check valves 1KBV-1243 through 1KVB-1250 [ sic] wc: Id provide a relatively leak tight boundary and ensure that the backup safet)-related pneumatic supplies for the chiller condenser cooling water pressure control valves would remain available for ]
four hours after a loss of power event.
(2) On April 8,1998, the backup safety-related pneumatic pressure regulators (1KBPCV-11464, -11466, and -11467) for the chiller condenser cooling water l
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Attachment '
LR-N980316 1
pressurscontrol valves were found set below minimum design requirements.
Operational tests had also not been performed to ensure that pressure regulators 1KBPCV-1164 [ sic) through 1KBPV-1171 [ sic] would remain property set in accordance with design requirements.
This is a Severity Level IV violation (Supplement 1). !
- 2. Reply to Notice of Violation PSE&G agrees with the violation. I
- 3. Reason for the Violation i Example (1)
PSE&G attributed the cause of the failure to establish inservice testing (IST) requirements and to periodically perform ISTs to personnel error. Personnel 1 preparing the design change package (DCP) and performing design specialty l reviews did not ensure the DCP was reviewed by the IST group. Deficiencies in the standard DCP format contributed to the violation. The design interface !
record had a single signoff for IST and for valve programs. I As part of the post-modification testing, an externalleak check was performed and the valves were functionally tested. Both tests were satisfactory. However, since the design change packages that added the check valves for the backup pneumatic supply were not reviewed by the IST group, IST requirements for the ;
valves were not established.
Example (2) ,
PSE&G attributed the cause of the failure to maintain the backup pneumatic pressure regulator settings to personnel error, most likely after the modification was completed. The regulator settings were verified as part of the post-modification testing. The most likely scenario is misadjustment by an operator during rounds or during a routine maintenance activity. Deficiencies in the procedure changes and in operator training for the design change package contributed to the violation.
I
- 4. Corrective Steps that Have Been Taken and Results Achieved Example (1)
- a. The check valves were tested satisfactorily.
- b. IST procedures for the check valves have been developed.
l c. The format for the design change interface record was revised to require a
{ separate signoff for the IST review.
Page 3 of 6 l
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Atta3hment LR-N980316
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- d. A detailed roll out of this event and its implications was provided to affected design engineering personnel. I
- e. Personnelinvolved were counseled conceming performance in this event.
Example (2)
- a. The backup pneumatic supply regulator settings were restored to their required values.
- b. Interim guidance on the operation of the backup pneumatic supply system was provided to operators. l l
l c. IST procedures for the check valves have been developed. The procedures include periodic verification of the pressure regulator settings for the backup pneumatic supply.
- 5. Corrective Steps that Will Be Taken to Avoid Further Violations Example (1) l No additional corrective actions are planned.
l 1
! Example (2)
- a. This violation will be incorporated in operator continuing training by September 1,1998.
- b. Lessons learned from this violation will be communicated to Engineering l personnel by September 30,1998.
i
- 6. Date When Full Compliance Will be Achieved l Example (1) l Hope Creek achieved full compliance on April 8,1998 when the inservice testing was performed satisfactorily on the backup pneumatic supply check valves. The i
valves have been added to the IST program.
I Example (2)
Hope Creek achieved fu'.I compliance on April 8,1998 when the backup pneumatic supply regulator settings were restored to their required values.
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Attachment LR-N980316 C.10 CFR 54Appe' ndix B Critorion XVI Violation
- 1. DesuW of the Notice of Violation 10 CFR Appendix B Criterion XVI (Corrective Action) requires, in part, that measures shall be established to assure that conditions adverse to quality, such i as failures, malfunctions, deficiencies, deviations, and nonconformances are promptly identified and corrected.
Contrary to the above, on December 10,1997, PSE&G engineers determined l that the minimum cooling water inlet temperature for the safety-related control area chilled water system chillers should be changed in a more limiting direction to 70 degrees Fahrenheit from 55 degrees Fehrenheit. On April 9,1998, the operations department management, still unaware of any necessary change to l the minimum allowed cooling water temperature, used 55 degrees Fahrenheit as t
a basis for determining inoperability when iney made a four-hour event notification to the NRC. Hope Creek abnormal operating procedure, Loss of Instrument Air and/or Service Air, HC.OP-AB.ZZ-0131(Q) - Rev.14, and pending change, HFSAR 97-080, to the Hope Creek Updated Final Safety Analysis i
Report (UFSAR) also incorrectly stated that 55 degrees Fahrenheit was the minimum cooling water temperature below which the safety-related backup pneumatic supply needed to remain operable. The change in minimum cooling l . water inlet temperature to a more limiting value was not corrected until May 7, i 1998, when guidance was provided to operators specifying the new 70 degrees Fahrenheit minimum cooling water temperature.
This is a Severity Level IV violation (Supplement 1).
- 2. Reply to Notice of Violation i
PSE&G agrees with the violation.
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- 3. Reason for the Violation PSE&G attributed the cause for this violation to personnel error. In December, 1997, the responsible engineer concluded that the minimum Safety Auxiliaries l Cooling system (SACS) temperature for Control Room chiller operation with full i
SACS flow is higher (more limiting) than the minimum temperature used for design of the backup pneumatic supply modification. The 55 degrees Fahrenheit temperature was an appropriate limit for a fully loaded chiller; but it is more conservative to assume that the chiller is lightly loaded. The responsible engineer, who is no longer employed by PSE&G, recognized the need for corrective action but did not initiate an Action Request as required by PSE&G's !
Corrective Action Program to ensure the non-conservative design assumption l was reviewed for its effect on chiller operability.
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Atta:hment LR-N980316
- 4. CorrectiVeTep's that Have B9en Taken and Results Achieved
- a. An Action Request to document this condition was initiated.
- b. The backup pneumatic supply was restored and the Control Room chillers were retumed to OPERABLE status on April 8,1998,
- c. A detailed roll out of this event and its implications was provided to affected design engineering personnel.
' 5. Corrective Steps that Will Be Taken to Avoid Further Violations
- a. An evaluation to determine the correct minimum SACS temperature for chiller operation without the Instrument Air system or backup pneumatic supply will be completed by August 21,1998. Temporary administrative controls are in i place to ensure the backup pneumatic supply remains in service when SACS temperature is less than 70 degrees Fahrenheit (for Control Room chillers) or 62 degrees Fahrenheit (for 1E Panel Room chiilers).
- b. Operating procedures will be revised as necessary by September 18,1998 to include the results of the evaluation described above.
- c. Lessons learned from this violation will be communicated to Engineering personnel by September 30,1998.
- 6. Date When Full Compliance Will be Achieved Hope Creek achieved full corripliance on April 8,1998 when the backup i
! pneumatic supply to the chiller pressure control valves was restored.
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