ML20212F995
| ML20212F995 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/29/1997 |
| From: | Graesser K COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-454-97-16, 50-455-97-16, BYRON-97-0242, BYRON-97-242, NUDOCS 9711050198 | |
| Download: ML20212F995 (6) | |
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nunnionmhh iaien ( unipan) n) run 6tnrrating Station i
61%O Mnth G(rtnan ( hun h Road ll);un, ll (i1010 9'9 6 hl NI% D 6 il 41 October 29,1997 LTR:
BYRON 97 0242 FILE:
1.10.0101 U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention:
Document Control Desk l
Subject:
Byron Nuclear Power Station Units 1 and 2 Response to Notice of Violation inspection Report No. 50-454/97016; 50-455/97016 NRC Docket Numbers 50-454,50-455 i
Reference:
John A. Grobe letter to Mr, Graesser dated September 30,1997, transmitting NRC Inspection Report 50-454/97016; 50-455/97016 Enclosed is Commenwealth Edison Company's response to the Notice of Violation (NOV) fA which was transmitted with the referenced letter and Inspection Report. The NOV cited one (1) Severity Level IV siolation requiring a written response. Comed's response is prosided in the attachment.
This letter contains the following commitments:
1)
An LER will be written in accordance with 10CFR50.73(a)(2)(v) & (si) documenting the concern 2)
The Reportability Manual will be revised to clarify the requirements for reportire srprocedural problems.
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Byron Ltr. 97 0242 October 29,1997
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1 Ifyour staf1'has any questions or comments concerning this letter, please refer them to
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Don Brindle, Regulatory Assurance Supervisor, at (815) 234 5441 ext. 2280.
9 Respectfully, i
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4N K. L. Grae er Site Vice President -
Byron Nuclear Power Station i
KLG/DB/rp Attachment (s)
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A. B. Beach, NRC Regional Administ:ntor - Rlli cc:
G. F. Dick Jr. Byron Project Manager - NRR Senior Resident inspector, Byron R. D. Lanksbury, Reactor Projects Chief-Rlli i
F. Niziolek, Division of Engineering - IDNS i
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ATTACllMENT I j
VIOLATION (454/455 07016 01) i Code of Federal Regulations Title 10 Part 50.72(b)(2)(iii) states, in part, that licensees shall notify the NRC when practical and in all cases, within four hours of the occurrence, "Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems needed to: (D) Mitigate the consequences of an accident."
Code of Federal Regulations Title 10 Part 50.73(a)(2)(v) states, in part, that licensees shall report, "Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to: (D) Mitigate the consequences of an accident."
Code of Federal Regulations Title 10 Part 50.73(a)(2)(vi) states that, " Events covered in paragraph (a)(2)(v) of this section may include one or more personnel errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies "
Contrary to the above, the inspector identified that the licensee on Febmary 19,1996, had failed to notify the NRC concerning procedural inadequacies with Byron emergency operating procedure (EOP) DEP-3, " Steam Generator Tube Rupture," and functional restoration procedure BFR P.1," Response to imminent Pressurized Thermal Shock Condition," which could limit operator response such that the EOP operator response time limits documented in the Updated Final Safety Analysis Report may not be met.
l This is a Severity Level IV violation (Supplement 1).
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lWASON F01C1111W10LAllDE The original concern with the EOPs was identified on Feb. 19,1996 in Problem identification Form (P!F)454 20196 0298 written by the EOP Procedure Writer. The concern was based on an Emergency Response Guideline (ERG) Direct Work Request
-(DW) DW 89-077 that had been submitted to the Westinghouse Owners Group (WOG) on November 10,1989 by Turkey Point. The EOP Procedure Writer had been reviewing the WOG response to the DW and felt that the response was inadequate. lie also felt that the issue pertained to llyron Station, and therefore wrote the PlF. The DW postulated that under some Steam Generator Tube Rupture (SGTR) scenarios it was possible that cold Safety injection (SI) flow could flow backwards through the Reactor Coolant System (RCS) loop and out the SG break if the Reactor Coolant Pumps (RCPs) were not running.
If this were to occur, it was further postulated that the cold Si flow might be such that the wide range RCS loop thermocouples would indicate a temperature lower than that used as criteria for entry into llyton Functional Response procedure llFR P.1, Response to imminent Pressurized Thermal Shock Condition." Ilyron Emergency Procedure llEP 3, Steam Generator Tube Rupture, did not provide any guidance to the operators that the indication oflow temperature in the ruptured RCS loop was expected and that Pressurized Thermal Shock (PTS) was not a concern. Thus, if the operators exited llEP 3 in order to follow IIRP P 1, it was then postulated that they might use up so much time responding to llFR P.1 that by the time they came back to 11EP-3, the SG would have overfilled before they had equalized RCS and SG pressures llyron's Design 13 asis does not include overfill of the SG's. Therefore, the issue was one of meeting the operator response time to SGTR assumed in the UFSAR. The DW response stated that "For most plants,it is expected that for a SGTR with RCPs tripped, the operator will remain in E 3 to properly respond to the SGTR."
The original PlF was reviewed for reportability by the Shill Manager. The Shift Manager wrote that the PlF was "Not an operability issue at this time pending further review.
Concern involves response in the emergency procedures and should be addressed by WOG." The Event Screening Committee also believed it was a generic WOG issue and assigned the PIF to the Emergency Procedure writer for resolution as a Level 4 (Apparent Cause Evaluation) PlF. The RA Supenisor consulted the Comed Reportability Manual.
Section SAF 1.17 of the manual addresses 10CFR50.72(b)(2)(iii) and 10CFR50.73(a)(2)(v) & (vi). Procedure problems are covered by the following paragraph:
A plant procedure, approved but not yet used, that has an error which would cause a safety system to become inoperable would be reportable. If the error was discovered before the procedure was approved, it would not be reportable.
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wm The root cause of the violation was the inappropriate decision made by the Regulatory Assurance Supenisor. The procedure problem did not "cause a safety system to become inoperable." It appeared that several circumstances would have had to occur in order to i
not meet the design basis, but none of them included making a system inoperable. The issue was also discussed with the RA Supenisor at Braidwood, who concurred with the Byron RA Supenisors position On 3/15/96, based on another review of the significance level, the PlFs significance level was raised to Level 3 (Root Cause Report), however it I
was still believed that the event was not reportable.
At that time, Byron Station was not aware of any other utilities that had made an NRC notification of the concern. Therefore, the Byron RA Supervisor made the decision that the issue was not reportable.
The Byron Emergency Procedure Writer was not satisfied with the WOG response to DW 89 077. On May 2,1996, he attended the WOG Operations Subcommittee meeting and again raised the concern with the BFR P.1 issue. The WOG Ops Subcommittee requested that he write another DW, which he did (DW 96 028). The DW was issued as Category 4 (Feedback to provide clarification or improve guidance (NOT to correct an error)). In early 1997 the WOG authorized a pregram to investigate operator action times On Feb. 28,1997, the WOG Ops Subcommittee issued a letter (OG 97 021) to applicable utilities asking them to nm selected time critical scenarios on the simulator, video taping them if possible, and to provide the results of the operator response back to l
the WOG. Byron and Braidwood both responded to the request. The WOG is reviewing the various responses and expects to issue a final report by the end of 1997. The second DW is still open.
CQERECTIVESTEPS TAKEN AED_RESULTS ACHIEVED in response to the originalissue with the BEP 3 and BFR P,1 procedures,it was realized 1
l that there might be other operator response times assumed in the UFSAR that had not been validated. Therefore, a Task Force was chartered to resiew those operator response times that were known at that time. The Task Force has reviewed 18 items since, with 16 items being closed out. Most of them have not been determined to be of concern.
Ilowever, several of them will be periodically validated to ensure that operators continue to meet the UFSAR assumed times.
As part of the Task Force response to the issue os'BEP-3 and DFR-P.1, a procedure change was made to BEP-3 that informed the operators that a low temperature condition was expected on the mptured loop, and that ifit occurred, to NOT go to DFR-P.l. If the operators do not go to DFR-P.1, they have time to equalize RCS and SG pressures prior to SG overfill.
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The RA Supenisor has been counseled on conservative decision making with respect to ENS noti 6 cations.
i The Byron RA Supenisor has discussed this issue with tne Braidwood RA Supenisor.
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CORRECTIVE STEPS TilAT WILL BE TAKEN TO AVOID FURTI1ER VIOLATION.
J An LER will be written in accordance with 10CFR50.73(a)(2)(v) & (vi) documenting the
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concern. This is tracked by NTS item # 454-100 97 01601-01.
The Reportability Manual will be revised to clarify the requirements for reporting of procedural problems. This is tracked by NTS item # 454 100 97 01601-02.
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DATE WilEN FULL C0htPLIANCE WILL BE ACHIEVED i
Full compliance will be achieved 30 days frorn the date of this letter when the LER will be submitted to the NRC.
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