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| See also: [[followed by::IR 05000244/1989017]]
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| =Text= | | =Text= |
| {{#Wiki_filter:ACCELERATED | | {{#Wiki_filter:ACCELERATED DISTRIBUTION DEMONST$&TION SYSTEM REGULATORY INFORMATION DISTRXBUTION SYSTEM (RIDS) |
| DISTRIBUTION | | ESSION NBR:9004040007 DOC ~ DATE: 90/03/26 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp. |
| DEMONST$&TION SYSTEM REGULATORY | | RECIP.NAME RECIPIENT, AFFILIATION RUSSELL,W.T; Region 1, Ofc of the Director R |
| INFORMATION | | |
| DISTRXBUTION | | ==SUBJECT:== |
| SYSTEM (RIDS)ESSION NBR:9004040007 | | Responds 50-244/89-17. |
| DOC~DATE: 90/03/26 NOTARIZED: | | to NRC 890222 ltr re violations noted in Insp Rept DISTRXBUTION CODE: IE01D COPIES RECEIVED:LTR ENCL 0 SIZE: |
| NO FACIL:50-244 | | TITLE: General (50 Dkt)-Insp Rept/Notice of Vi lation Response, NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).. 05000244,'] |
| Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION | | RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 PD JOHNSON,A INTERNAL'EOD 1 AEOD/DEIIB 1 AEOD/TPAD 1 DEDRO 1 NRR SHANKMAN,S 1 NRR/DLPQ/LPEB10 1 NRR/DOEA DIR 11 1 NRR/DREP/PEPB9D 1 NRR/DREP/PRPB11 ,2 |
| MECREDY,R.C. | | '1 NRR/DRIS/DIR 1 NRR/DST/DXR 8E2 NRR/PMAS/ILRB12 l NUDOCS=ABSTRACZ 1 OGC/HDS2 1 REG FIXE'- ~02~ 1 RES MORISSEAU,D 1 RGN1 FILE 01 1 EXTERNAL: LPDR 1 NRC PDR NSIC 1 legs p]5 7~ '-' |
| Rochester Gas&Electric Corp.RECIP.NAME | | .A NOTE TO ALL"RIDS" RECIPIENTS: |
| RECIPIENT, AFFILIATION | | PLEASE HELP US TO REDUCE WAS'ONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED! |
| RUSSELL,W.T; | | OTAL NUMBER OF COPIES REQUIRED: LTTR 23 ENCL |
| Region 1, Ofc of the Director SUBJECT: Responds to NRC 890222 ltr re violations | | |
| noted in Insp Rept 50-244/89-17. | | I |
| DISTRXBUTION
| | ~ |
| CODE: IE01D COPIES RECEIVED:LTR | | |
| ENCL 0 SIZE: TITLE: General (50 Dkt)-Insp Rept/Notice | | f f A'f f~ ff ff RTC If f,i i 'TAN I |
| of Vi lation Response, DOCKET 05000244 R NOTES:License | | ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER, N.Y. 14849-pppg March 26, 1990 TCKCRHONC ARCA COOK 71K 546 2700 Mr. William T. Russell Regional Administrator U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406 |
| Exp date in accordance | | |
| with 10CFR2,2.109(9/19/72).. | | ==Subject:== |
| 05000244,'] | | Response to Notices of Violation Inspection Report No. 50-244/89-17 R.E. Ginna Nuclear Power Plant Docket No. 50-244 |
| RECIPIENT ID CODE/NAME PD1-3 PD INTERNAL'EOD | | |
| AEOD/TPAD NRR SHANKMAN,S | | ==Dear Mr. Russell:== |
| NRR/DOEA DIR 11 NRR/DREP/PRPB11 | | |
| NRR/DST/DXR | | This letter is in response to the February 22, 1989 letter from Jon R. Johnson, Chief, Projects Branch No. 3 to Robert E. Smith, Senior Vice President, RG&E, which transmitted Inspection Report No. 50-244/89-17. In that report, two violations were identified. The following provides a reply to the violations pursuant to 10 CFR 2.201. |
| 8E2 NUDOCS=ABSTRACZ | | RESTATEMENT OF VIOLATIONS During inspection at the R.E. Ginna Nuclear Power Plant from December 12, 1989 through January 8, 1990, the following violations were identified and evaluated in accordance with the NRC Enforcement Policy (10 CFR 2, Appendix C): |
| REG FIXE'--~02~RGN1 FILE 01 EXTERNAL: LPDR NSIC COPIES LTTR ENCL 1 1 1 1 ,2'1 1 1 1 1 1 RECIPIENT ID CODE/NAME JOHNSON,A AEOD/DEIIB | | A. 10 CFR 50, Appendix B, Criterion XVI, and the Ginna Quality Assurance Manual, Section 16, require prompt identification and correction of conditions adverse to quality including failures, malfunctions, deficiencies, defective material and equipment, and nonconformances. |
| DEDRO NRR/DLPQ/LPEB10
| | Contrary to the above, a safety injection system design deficiency was not promptly identified and corrected when corporate engineering was notified on or before October 20, '989 that failure of the safety injection block/unblock switch could block automatic safety injection actuation on low pressurizer pressure or low steam line pressure. Corporate engineering did. not conclude that this problem existed at Ginna until about November 17, 1989, and site technical personnel were not informed about the deficiency until December 19, 1989. |
| NRR/DREP/PEPB9D
| | This is a Severity Level IV violation (Supplement I). |
| NRR/DRIS/DIR
| | /0040 ">0V07 200 ADOCI''=000:..44 c'OR FDC ~ |
| NRR/PMAS/ILRB12
| | ~ Qo |
| OGC/HDS2 RES MORISSEAU,D
| | ~l" |
| NRC PDR COPIES LTTR ENCL 1 1 1 1 1 l 1 1 legs p]5 7~'-'.A NOTE TO ALL"RIDS" RECIPIENTS:
| | |
| PLEASE HELP US TO REDUCE WAS'ONTACT
| | 4 B. 10 CFR 50, Appendix B, Criterion V, and the Ginna Quality Assurance Manual, Section 5, require activities affecting quality to be accomplished in accordance with instructions, procedures, or drawings which include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. |
| THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROM DISTRIBUTION
| | Contrary to the above, on December 15, 1989, maintenance was performed on a safety-related motor-operated valve in the safety injection system in accordance with a procedure which included an inappropriate torque specification. |
| LISTS FOR DOCUMENTS YOU DON'T NEED!OTAL NUMBER OF COPIES REQUIRED: LTTR 23 ENCL
| | This is a Severity Level V violation (Supplement I). |
| I~
| | RESPONSE TO VIOLATION A RG&E Position on Existence of Violation Rochester Gas and Electric Corporation (RG&E) concurs that a violation of Appendix B, Criterion XVI occurred. RG&E recognizes that communication between corporate engineering and site personnel on issues of potential safety significance should be formalized. Our efforts to address this concern are provided in Section 4, "Long Term Enhancements". As explained below, RG&E also believes that with respect to the issue identified on October 20, 1989, we acted in a manner consistent with the safety |
| ROCHESTER GAS f f A'f f~ff ff RTC If f,i i'TAN I AND ELECTRIC CORPORATION
| | .significance of the matter. |
| ~89 EAST AVENUE, ROCHESTER, N.Y.14849-pppg
| | : 2. Reason for Violation As Inspection Report No. 50-244/89-17 (p. 7) indicates, RG&E received notice on October 20, 1989, from Westinghouse Electric Corporation (Westinghouse) of an apparent generic design deficiency related to the type of safety injection (SI) block/unblock switch used at various Westinghouse reactors. The Westinghouse letter, dated October 12, 1989, concluded that a "single failure of the switch (Westinghouse OT2) could block either the automatic low pressurizer pressure or the low steamline pressure SI signal in both trains" [emphasis supplied]. The letter also' stated that the probability of switch failure was "10 10 '/yr" :and that, while a design change was recommended, the situation was "not an immediate safety concern." |
| March 26, 1990 TCKCRHONC ARCA COOK 71K 546 2700 Mr.William T.Russell Regional Administrator
| | In addition, the Westinghouse letter referred to a Licensee Event Report (LER), No. 88-007-00, submitted by Wisconsin, Electric Power Company (Wisconsin Electric) on September 16, 1988, concerning the same issue at the Point Beach Nuclear Plant (Point Beach). The Wisconsin Electric LER concluded that "this condition will not have a significant impact on the health and safety of the general public or the employees of the Point Beach Nuclear Plant." |
| U.S.Nuclear Regulatory
| | |
| Commission
| | The LER noted that the Point Beach facility was operating at 100% capacity when the concern was identified and that design change would not'e made until the next scheduled outage. |
| Region I 475 Allendale Road King of Prussia, Pennsylvania
| | Upon receipt of the Westinghouse notification on October 20, 1989, RG&E (corporate) initiated a timely review for applicability to Ginna Station. Based on the Wisconsin Electric LER and on Westinghouse's calculation of the low probability of switch failure, it was apparent that the matter did not constitute an immediate safety concern. |
| 19406 Subject: Response to Notices of Violation Inspection
| | When it was identified that the switch configuration was applicable to Ginna Station, an internal engineering recommendation was made consistent with the guidance of the Westinghouse letter and attached LER, that an EWR be initiated. This was completed on November 17, 1989. This recommendation was then evaluated within Nuclear Safety and Licensing, resulting in a discussion with site technical support personnel relative to this situation on December 19, 1989. On December 20, site personnel initiated a Ginna Station Event Report per Procedure A-25.1 (Event No. 89-168). The event report indicated that the site Plant Operations Review Committee (PORC) had, on December 20, 1989, concluded that plant operation could continue for the following reasons: |
| Report No.50-244/89-17
| | very low (i.e., 10 'o |
| R.E.Ginna Nuclear Power Plant Docket No.50-244 Dear Mr.Russell: This letter is in response to the February 22, 1989 letter from Jon R.Johnson, Chief, Projects Branch No.3 to Robert E.Smith, Senior Vice President, RG&E, which transmitted
| | : 1. Westinghouse stated that the. probability of failure was 10 '/yr); |
| Inspection
| | : 2. Emergency Operating Procedures directed Operators to use manual SI initiation where indicators show automatic initiation has failed; |
| Report No.50-244/89-17.
| | : 3. A separate automatic SI initiating mechanism would activate when containment pressure reached 4 psig; |
| In that report, two violations
| | : 4. During depressurization, a bistable light will'lert operators of a blocked SI signal; and |
| were identified. | | : 5. Visual verification of the SI switch plunger position indicates that the contacts are in the proper position. |
| The following provides a reply to the violations
| | The violation states that the time between October 20, 1989, when RG&E (corporate) was notified by Westinghouse, and the communication of this information to the site technical staff on December 19, 1989, shows that the SI design deficiency was not promptly identified and corrected, and indicates problems in communication between corporate engineering and site personnel. While RG&E does not deny this violation, we believe that the actions taken by RG&E were appropriate in view of RG&E's preliminary conclusion that the issue did not constitute an immediate safety concern. |
| pursuant to 10 CFR 2.201.RESTATEMENT
| | |
| OF VIOLATIONS
| | RG&E believes that Appendix B, Criterion XVI does not establish a precise time limit for resolution of safety issues. Rather, issues such as "promptness" or "timeliness" are subjective matters that inherently depend upon the safety significance of the situation. Given that RGGE had a documented recommendation from Westinghouse that no immediate safety concern existed (as corroborated by the Point Beach LER), its actions toward resolution of the issue were prompt and timely. Any other interpretations of Criterion XVI would be counter to public health and safety because it would require licensees to treat all deficiencies or non-conforming items the same (i.e., regardless of safety significance). |
| During inspection
| | This same basic philosophy was affirmed in an analogous context 'in recent guidance issued by NRC's Office of Nuclear Reactor Regulation '(NRR). Specifically, on July 19, 1989, Dr. T.E. Murley, Director, NRC/NRR, sent a memorandum to all of the regional administrators entitled "Guidance on Action To Be Taken Following Discovery of Potentially Nonconforming Equipment." In his memorandum, Dr. Murley stated that "[t]here is no generally appropriate timeframe in which operability determinations should be made." For equipment which is "clearly inoperable," an immediate declaration of inoperability should be made and the appropriate technical specifications followed. However, Dr. Murley's memorandum contrasts this situation with those where equipment nonconformances simply raise the issue of operability. |
| at the R.E.Ginna Nuclear Power Plant from December 12, 1989 through January 8, 1990, the following violations
| | In such situations Dr. Murley states that: |
| were identified
| | operability determinations should be made by licensees as soon as racticable, and in a timeframe commensurate with the a licable e ui ment's im ortance to safet usin the best information available,(e.g., analyses, a test or partial test, experience with operating events, engineering judgement or a combination of the factors) (emphasis supplied). |
| and evaluated in accordance | | Although this guidance relates to timing of operability determinations, it is equally appropriate with respect to resolution of open items under Criterion XVI. Consistent with this philosophy and based on the best information available, future cases of this type will be resolved "as soon as practicable" and in a time commensurate with the safety significance of the matter. Communication between corporate and site personnel will be initiated promptly once applicability to Ginna Station is determined. |
| with the NRC Enforcement
| | Corrective Ste s Which Have Been Taken and the Results Achieved Corporate and site technical staff and the PORC have reviewed the circumstances surrounding the potentially generic design deficiency related to the control room SI block/unblock switch. As stated in LER 89-016, the. |
| Policy (10 CFR 2, Appendix C): Contrary to the above, a safety injection system design deficiency
| | following actions were taken: |
| was not promptly identified | | |
| and corrected when corporate engineering | | Knowledgeable personnel inspected the plunger position of the SI Block/Unblock Switch and verified that the switch contacts were in the proper position. |
| was notified on or before October 20,'989 that failure of the safety injection block/unblock | | ~ Operating Procedure 0-1.1 (Plant Heatup From Cold Shutdown to Hot Shutdown) was changed to add the following note and check-off to Step 5.11.6: |
| switch could block automatic safety injection actuation on low pressurizer | | NOTE: Prior to placing the SI Block/Unblock Switch to the normal position, station an operator inside the MCB in direct observation of the SI Block/Unblock Switch to observe that both plunger tips are recessed inward after the switch is placed. to normal position.- |
| pressure or low steam line pressure.Corporate engineering | | Block switch plunger t'ips position inward |
| did.not conclude that this problem existed at Ginna until about November 17, 1989, and site technical personnel were not informed about the deficiency | | ~ An RG&E operator aid tag was placed on the .MCB adjacent to the SI Block/Unblock Switch denoting the note- from 0-1.1. |
| until December 19, 1989.This is a Severity Level IV violation (Supplement | | ~ An RG&E operator aid tag was also placed inside the MCB adj acent to the rear of the SI Block/Unblock Switch stating the following: This is the switch we verify that the plunger's tips are recessed inward when the switch is placed to normal (labeled LAK) . |
| I).~Qo~~l"/0040">0V07 200 c'OR ADOCI''=000:..44 | | A spare switch of similar design has been placed in the Control Room for the purpose of training the operators to recognize the differences in plunger position. |
| FDC A.10 CFR 50, Appendix B, Criterion XVI, and the Ginna Quality Assurance Manual, Section 16, require prompt identification
| | These actions are considered adequate to provide reasonable assurance of SI system operability until the situation can be permanently dispositioned. Finally, EWR 5025 was initiated to provide for the installation of independent SI block/unblock switches for each SI train which is planned for the 1991 refueling outage. |
| and correction
| | : 4. Corrective Ste s Which Will Be Taken to Avoid Further Violation RG&E has recently taken steps to upgrade the overall corrective action program for Ginna Station. The need for improvements was noted during the course of the RHR System Safety System Functional Inspection (SSFI), and is also considered appropriate due to RG&E's initiation of a comprehensive Configuration Management/Design Basis Program. We are working with the NUMARC Design Basis Issues Working Group to develop an improved problem identification and resolution program. |
| of conditions
| | The improved program will: |
| adverse to quality including failures, malfunctions, deficiencies, defective material and equipment, and nonconformances.
| | ~ Improve the process of identifying, analyzing, and resolving problems; |
| 4
| | |
| B.10 CFR 50, Appendix B, Criterion V, and the Ginna Quality Assurance Manual, Section 5, require activities
| | ~ Improve the RG&E internal review process, including formalized means of communication between corporate engineering and site personnel on issues of potential safety significance; and Part of the implementation of this effort will include specific procedural upgrades, enhancement of our corrective action tracking system, and the issuance of a corporate policy which addresses problem identification and reporting. We believe that this broad effort, when fully implemented, will improve our capability to consistently identify and disposition potential safety issues commensurate with their significance. |
| affecting quality-to be accomplished
| | : 5. Date When Full Com liance Will Be Achieved Long term and short term actions and schedules have been described above. Formal guidance concerning communication between corporate and site personnel on identified problem issues is under development, and is targeted for completion by July 1990. |
| in accordance | | RESPONSE TO VIOLATION B Rochester Gas and Electric concurs with this violation as stated below. |
| with instructions, procedures, or drawings which include appropriate | | Reason for Violation Rochester Gas and Electric agrees that, Ginna Station does not have an established written policy regarding consideration of inherent inaccuracy of calibrated measuring and test, equipment (M&TE) when developing acceptance criteria. |
| quantitative
| | As- a common practice, torquing methods address only instru-ment "indication" and are not meant to include the instrument accuracy. This practice is based on the fact that torque is only a general indicator of bolting pre-load because of the inaccuracies, e.g., lubrication, thread fit, thread condition, etc., inherent in the torque equation. When highly accurate bolt pre-loading is required, means other than torque is used, i.e., stud elongation to determine bolt pre-load. |
| or qualitative
| | The Corrective Ste s Which Have Been Taken and the Results Achieved Due to the successful completion of post maintenance testing, no action regarding the valve packing adjustment has been taken. |
| acceptance
| | A-1603.4, "Work Order Scheduling" was revised to require work and testing to be completed on individual trains prior to starting maintenance on a redundant train. |
| criteria for determining
| | |
| that important activities | | ' The Corrective Ste s Which Will Be Taken to Avoid Further Violation |
| have been satisfactorily
| | : 1. Administrative procedure A-1603.3, "Work Order Planning" will be revised to state a Ginna Station policy regarding consideration of M&TE inherent inaccuracy and provide direction for development'f acceptance criteria utilizi'ng this equipment. |
| accomplished.
| | : 2. A new procedure for packing adjustment is being developed to provide specific direction for adjustment of valves repacked under the Valve Packing Improvement Program and to provide a method of maintaining and updating valve packing data. |
| Contrary to the above, on December 15, 1989, maintenance
| | The Date When Full Com liance Will Be Achieved The anticipated effective date of the above procedures is May 1, 1990, for the maintenance procedures and June 30, 1990, for the administrative procedure. |
| was performed on a safety-related
| | Very truly yours, Robert C. Me dy Division Manager Nuclear Production GJWN093 Enclosures xc: U.S. Nuclear Regulatory Commission (original) |
| motor-operated
| | Document Control Desk Washington, D.C. 20555 Allen R. Johnson .(Mail Stop 14D1) |
| valve in the safety injection system in accordance
| | Project Directorate I-3 Washington, D.C. 20555 Nicholas S. Reynolds, Esq. |
| with a procedure which included an inappropriate
| | Bishop, Cook, Purcell and Reynolds 1400 L. Street, N.W. |
| torque specification.
| | Washington, D.C. 20005-3502 Ginna NRC Senior Resident Inspector |
| This is a Severity Level V violation (Supplement
| | |
| I).RESPONSE TO VIOLATION A RG&E Position on Existence of Violation Rochester Gas and Electric Corporation (RG&E)concurs that a violation of Appendix B, Criterion XVI occurred.RG&E recognizes
| | ~l 0}} |
| that communication
| |
| between corporate engineering
| |
| and site personnel on issues of potential safety significance
| |
| should be formalized.
| |
| Our efforts to address this concern are provided in Section 4,"Long Term Enhancements".
| |
| As explained below, RG&E also believes that with respect to the issue identified
| |
| on October 20, 1989, we acted in a manner consistent
| |
| with the safety.significance
| |
| of the matter.2.Reason for Violation As Inspection
| |
| Report No.50-244/89-17 (p.7)indicates, RG&E received notice on October 20, 1989, from Westinghouse
| |
| Electric Corporation (Westinghouse)
| |
| of an apparent generic design deficiency
| |
| related to the type of safety injection (SI)block/unblock
| |
| switch used at various Westinghouse
| |
| reactors.The Westinghouse
| |
| letter, dated October 12, 1989, concluded that a"single failure of the switch (Westinghouse
| |
| OT2)could block either the automatic low pressurizer
| |
| pressure or the low steamline pressure SI signal in both trains"[emphasis supplied].
| |
| The letter also stated that the probability
| |
| of switch failure was"10'10'/yr":and that, while a design change was recommended, the situation was"not an immediate safety concern." In addition, the Westinghouse
| |
| letter referred to a Licensee Event Report (LER), No.88-007-00, submitted by Wisconsin, Electric Power Company (Wisconsin
| |
| Electric)on September 16, 1988, concerning
| |
| the same issue at the Point Beach Nuclear Plant (Point Beach).The Wisconsin Electric LER concluded that"this condition will not have a significant
| |
| impact on the health and safety of the general public or the employees of the Point Beach Nuclear Plant."
| |
| '
| |
| The LER noted that the Point Beach facility was operating at 100%capacity when the concern was identified
| |
| and that design change would not'e made until the next scheduled outage.Upon receipt of the Westinghouse
| |
| notification
| |
| on October 20, 1989, RG&E (corporate)
| |
| initiated a timely review for applicability
| |
| to Ginna Station.Based on the Wisconsin Electric LER and on Westinghouse's | |
| calculation
| |
| of the low probability
| |
| of switch failure, it was apparent that the matter did not constitute | |
| an immediate safety concern.When it was identified
| |
| that the switch configuration | |
| was applicable
| |
| to Ginna Station, an internal engineering
| |
| recommendation
| |
| was made consistent | |
| with the guidance of the Westinghouse
| |
| letter and attached LER, that an EWR be initiated.
| |
| This was completed on November 17, 1989.This recommendation
| |
| was then evaluated within Nuclear Safety and Licensing, resulting in a discussion
| |
| with site technical support personnel relative to this situation on December 19, 1989.On December 20, site personnel initiated a Ginna Station Event Report per Procedure A-25.1 (Event No.89-168).The event report indicated that the site Plant Operations
| |
| Review Committee (PORC)had, on December 20, 1989, concluded that plant operation could continue for the following reasons: 1.Westinghouse
| |
| stated that the.probability
| |
| of failure was very low (i.e., 10'o 10'/yr);2.Emergency Operating Procedures
| |
| directed Operators to use manual SI initiation
| |
| where indicators
| |
| show automatic initiation
| |
| has failed;3.A separate automatic SI initiating
| |
| mechanism would activate when containment
| |
| pressure reached 4 psig;4.During depressurization, a bistable light will'lert operators of a blocked SI signal;and 5.Visual verification
| |
| of the SI switch plunger position indicates that the contacts are in the proper position.The violation states that the time between October 20, 1989, when RG&E (corporate)
| |
| was notified by Westinghouse, and the communication
| |
| of this information | |
| to the site technical staff on December 19, 1989, shows that the SI design deficiency
| |
| was not promptly identified
| |
| and corrected, and indicates problems in communication
| |
| between corporate engineering
| |
| and site personnel.
| |
| While RG&E does not deny this violation, we believe that the actions taken by RG&E were appropriate
| |
| in view of RG&E's preliminary
| |
| conclusion
| |
| that the issue did not constitute | |
| an immediate safety concern.
| |
|
| |
| RG&E believes that Appendix B, Criterion XVI does not establish a precise time limit for resolution
| |
| of safety issues.Rather, issues such as"promptness" or"timeliness" are subjective
| |
| matters that inherently
| |
| depend upon the safety significance
| |
| of the situation. | |
| Given that RGGE had a documented
| |
| recommendation
| |
| from Westinghouse
| |
| that no immediate safety concern existed (as corroborated
| |
| by the Point Beach LER), its actions toward resolution
| |
| of the issue were prompt and timely.Any other interpretations
| |
| of Criterion XVI would be counter to public health and safety because it would require licensees to treat all deficiencies | |
| or non-conforming
| |
| items the same (i.e., regardless
| |
| of safety significance). | |
| This same basic philosophy
| |
| was affirmed in an analogous context'in recent guidance issued by NRC's Office of Nuclear Reactor Regulation
| |
| '(NRR).Specifically, on July 19, 1989, Dr.T.E.Murley, Director, NRC/NRR, sent a memorandum
| |
| to all of the regional administrators
| |
| entitled"Guidance on Action To Be Taken Following Discovery of Potentially
| |
| Nonconforming
| |
| Equipment." In his memorandum, Dr.Murley stated that"[t]here is no generally appropriate
| |
| timeframe in which operability
| |
| determinations
| |
| should be made." For equipment which is"clearly inoperable," an immediate declaration
| |
| of inoperability | |
| should be made and the appropriate
| |
| technical specifications
| |
| followed.However, Dr.Murley's memorandum
| |
| contrasts this situation with those where equipment nonconformances
| |
| simply raise the issue of operability.
| |
| In such situations
| |
| Dr.Murley states that: operability
| |
| determinations
| |
| should be made by licensees as soon as racticable, and in a timeframe commensurate
| |
| with the a licable e ui ment's im ortance to safet usin the best information
| |
| available,(e.g., analyses, a test or partial test, experience
| |
| with operating events, engineering
| |
| judgement or a combination
| |
| of the factors)(emphasis supplied).
| |
| Although this guidance relates to timing of operability
| |
| determinations, it is equally appropriate
| |
| with respect to resolution
| |
| of open items under Criterion XVI.Consistent | |
| with this philosophy
| |
| and based on the best information
| |
| available, future cases of this type will be resolved"as soon as practicable" and in a time commensurate
| |
| with the safety significance
| |
| of the matter.Communication | |
| between corporate and site personnel will be initiated promptly once applicability
| |
| to Ginna Station is determined. | |
| Corrective
| |
| Ste s Which Have Been Taken and the Results Achieved Corporate and site technical staff and the PORC have reviewed the circumstances
| |
| surrounding
| |
| the potentially
| |
| generic design deficiency
| |
| related to the control room SI block/unblock
| |
| switch.As stated in LER 89-016, the.following actions were taken:
| |
|
| |
| Knowledgeable
| |
| personnel inspected the plunger position of the SI Block/Unblock
| |
| Switch and verified that theswitch contacts were in the proper position.~Operating Procedure 0-1.1 (Plant Heatup From Cold Shutdown to Hot Shutdown)was changed to add the following note and check-off to Step 5.11.6: NOTE: Prior to placing the SI Block/Unblock
| |
| Switch to the normal position, station an operator inside the MCB in direct observation
| |
| of the SI Block/Unblock
| |
| Switch to observe that both plunger tips are recessed inward after the switch is placed.to normal position.-
| |
| Block switch plunger t'ips position inward~An RG&E operator aid tag was.placed on the.MCB adjacent to the SI Block/Unblock
| |
| Switch denoting the note-from 0-1.1.~An RG&E operator aid tag was also placed inside the MCB adj acent to the rear of the SI Block/Unblock
| |
| Switch stating the following:
| |
| This is the switch we verify that the plunger's tips are recessed inward when the switch is placed to normal (labeled LAK).A spare switch of similar design has been placed in the Control Room for the purpose of training the operators to recognize the differences
| |
| in plunger position.These actions are considered
| |
| adequate to provide reasonable
| |
| assurance of SI system operability
| |
| until the situation can be permanently
| |
| dispositioned.
| |
| Finally, EWR 5025 was initiated to provide for the installation
| |
| of independent | |
| SI block/unblock
| |
| switches for each SI train which is planned for the 1991 refueling outage.4.Corrective
| |
| Ste s Which Will Be Taken to Avoid Further Violation RG&E has recently taken steps to upgrade the overall corrective
| |
| action program for Ginna Station.The need for improvements
| |
| was noted during the course of the RHR System Safety System Functional
| |
| Inspection (SSFI), and is also considered
| |
| appropriate
| |
| due to RG&E's initiation
| |
| of a comprehensive
| |
| Configuration
| |
| Management/Design
| |
| Basis Program.We are working with the NUMARC Design Basis Issues Working Group to develop an improved problem identification
| |
| and resolution
| |
| program.The improved program will:~Improve the process of identifying, analyzing, and resolving problems;
| |
|
| |
| ~Improve the RG&E internal review process, including formalized
| |
| means of communication
| |
| between corporate engineering
| |
| and site personnel on issues of potential safety significance;
| |
| and Part of the implementation | |
| of this effort will include specific procedural
| |
| upgrades, enhancement
| |
| of our corrective | |
| action tracking system, and the issuance of a corporate policy which addresses problem identification
| |
| and reporting. | |
| We believe that this broad effort, when fully implemented, will improve our capability
| |
| to consistently
| |
| identify and disposition
| |
| potential safety issues commensurate
| |
| with their significance.
| |
| 5.Date When Full Com liance Will Be Achieved Long term and short term actions and schedules have been described above.Formal guidance concerning
| |
| communication
| |
| between corporate and site personnel on identified
| |
| problem issues is under development, and is targeted for completion
| |
| by July 1990.RESPONSE TO VIOLATION B Rochester Gas and Electric concurs with this violation as stated below.Reason for Violation Rochester Gas and Electric agrees that, Ginna Station does not have an established
| |
| written policy regarding consideration
| |
| of inherent inaccuracy
| |
| of calibrated
| |
| measuring and test, equipment (M&TE)when developing
| |
| acceptance
| |
| criteria.As-a common practice, torquing methods address only instru-ment"indication" and are not meant to include the instrument
| |
| accuracy.This practice is based on the fact that torque is only a general indicator of bolting pre-load because of the inaccuracies, e.g., lubrication, thread fit, thread condition, etc., inherent in the torque equation.When highly accurate bolt pre-loading
| |
| is required, means other than torque is used, i.e., stud elongation
| |
| to determine bolt pre-load.The Corrective
| |
| Ste s Which Have Been Taken and the Results Achieved Due to the successful
| |
| completion
| |
| of post maintenance
| |
| testing, no action regarding the valve packing adjustment
| |
| has been taken.A-1603.4,"Work Order Scheduling" was revised to require work and testing to be completed on individual
| |
| trains prior to starting maintenance
| |
| on a redundant train.
| |
|
| |
| 'The Corrective
| |
| Ste s Which Will Be Taken to Avoid Further Violation 1.Administrative
| |
| procedure A-1603.3,"Work Order Planning" will be revised to state a Ginna Station policy regarding consideration
| |
| of M&TE inherent inaccuracy
| |
| and provide direction for development'f
| |
| acceptance
| |
| criteria utilizi'ng
| |
| this equipment.
| |
| 2.A new procedure for packing adjustment
| |
| is being developed to provide specific direction for adjustment
| |
| of valves repacked under the Valve Packing Improvement
| |
| Program and to provide a method of maintaining
| |
| and updating valve packing data.The Date When Full Com liance Will Be Achieved The anticipated
| |
| effective date of the above procedures
| |
| is May 1, 1990, for the maintenance
| |
| procedures
| |
| and June 30, 1990, for the administrative
| |
| procedure.
| |
| Very truly yours,Robert C.Me dy Division Manager Nuclear Production
| |
| GJWN093 Enclosures
| |
| xc: U.S.Nuclear Regulatory
| |
| Commission (original)
| |
| Document Control Desk Washington, D.C.20555 Allen R.Johnson.(Mail Stop 14D1)Project Directorate
| |
| I-3 Washington, D.C.20555 Nicholas S.Reynolds, Esq.Bishop, Cook, Purcell and Reynolds 1400 L.Street, N.W.Washington, D.C.20005-3502
| |
| Ginna NRC Senior Resident Inspector
| |
| ~l 0 | |
| }} | |
Similar Documents at Ginna |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated IR 05000244/19990081999-10-14014 October 1999 Forwards Insp Rept 50-244/99-08 on 990809-0919.Severity Level IV Violation of NRC Requirements Occurred & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs IR 05000244/19992011999-09-24024 September 1999 Forwards Insp Rept 50-244/99-201 (Operational Safeguards Response Evaluation) on 990621-24.No Violations Noted. Primary Purpose of Osre to Assess Licensee Ability to Respond to External Threat.Insp Rept Withheld ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification ML17309A6551999-06-0707 June 1999 Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML17265A6671999-06-0101 June 1999 Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301 ML17265A6641999-05-25025 May 1999 Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested ML17265A6611999-05-24024 May 1999 Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730 ML20206U0921999-05-13013 May 1999 Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21 ML17265A6461999-05-12012 May 1999 Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP ML17265A6411999-05-12012 May 1999 Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630 ML20206E7221999-04-29029 April 1999 Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21 ML20206H6911999-04-22022 April 1999 Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls ML17265A6271999-04-19019 April 1999 Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5 ML17309A6501999-04-14014 April 1999 Forwards Revised Ginna Station EOPs & Procedures Index ML17265A6121999-03-29029 March 1999 Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231 ML17265A6041999-03-24024 March 1999 Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618 ML17265A5671999-03-0101 March 1999 Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl ML17265A5641999-03-0101 March 1999 Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl ML17265A5551999-02-25025 February 1999 Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr ML17265A5561999-02-22022 February 1999 Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d) ML17265A5451999-02-12012 February 1999 Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii) ML17309A6491999-02-12012 February 1999 Forwards Ginna Station EOPs ML17265A5431999-02-0909 February 1999 Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301 ML17265A5361999-02-0202 February 1999 Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl ML17311A0691999-01-25025 January 1999 Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date ML17311A0671999-01-14014 January 1999 Forwards Revised Emergency Operating Procedures for Re Ginna NPP ML17265A5131999-01-12012 January 1999 Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification ML17309A6551999-06-0707 June 1999 Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML17265A6671999-06-0101 June 1999 Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301 ML17265A6641999-05-25025 May 1999 Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested ML17265A6611999-05-24024 May 1999 Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730 ML20206U0921999-05-13013 May 1999 Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21 ML17265A6461999-05-12012 May 1999 Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP ML17265A6411999-05-12012 May 1999 Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630 ML20206E7221999-04-29029 April 1999 Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21 ML20206H6911999-04-22022 April 1999 Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls ML17265A6271999-04-19019 April 1999 Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5 ML17309A6501999-04-14014 April 1999 Forwards Revised Ginna Station EOPs & Procedures Index ML17265A6121999-03-29029 March 1999 Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231 ML17265A6041999-03-24024 March 1999 Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618 ML17265A5641999-03-0101 March 1999 Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl ML17265A5671999-03-0101 March 1999 Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl ML17265A5551999-02-25025 February 1999 Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr ML17265A5561999-02-22022 February 1999 Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d) ML17265A5451999-02-12012 February 1999 Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii) ML17309A6491999-02-12012 February 1999 Forwards Ginna Station EOPs ML17265A5431999-02-0909 February 1999 Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301 ML17265A5361999-02-0202 February 1999 Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl ML17311A0691999-01-25025 January 1999 Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date ML17311A0671999-01-14014 January 1999 Forwards Revised Emergency Operating Procedures for Re Ginna NPP ML17265A5131999-01-12012 January 1999 Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date ML17265A5111999-01-11011 January 1999 Requests Relief Per 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME Bp&V Code for ISI Program.Relief Requests 37,38 & 39 Encl ML17265A5101999-01-11011 January 1999 Requests Relief Per to 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 40 Encl 1999-09-30
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L4481990-09-18018 September 1990 Forwards Responses to Rer 900709-13 Team Visit Findings, Per .Responses Withheld (Ref 10CFR73.21) ML17309A4491990-09-13013 September 1990 Forwards Slides Presented by Facility & Westinghouse to NRC in 900724 Meeting.Encl Withheld ML17262A1331990-09-11011 September 1990 Responds to Violations & Several Unresolved Items Noted in SSFI Rept 50-244/89-81.Update of Appropriate Unresolved Items Encl.Specific Actions Re All NRC Unresolved Items Being Tracked to Completion ML17261B1511990-08-29029 August 1990 Forwards Semiannual Radioactive Effluent Release Rept Jan- June 1990 & Rev 3 to Process Control Program for Ginna Station, Per Tech Specs 6.9.1.4 & 6.16,respectively ML17261B1481990-08-28028 August 1990 Lists Understanding of Issues Util Planning to Address Re Containment Integrity,Per 900718 Telcon.Any Concerns or Action Items Different from Listed Submittal Should Be Provided to Util Prior to NRC 900905 & 06 Visit to Plant IR 05000244/19880261990-08-20020 August 1990 Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented ML17261B1501990-08-20020 August 1990 Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented ML17261B1371990-08-17017 August 1990 Forwards Rev 0 to, Inservice Insp Rept for Third Interval (1990-1999),First Period,First Outage (1990) at Re Ginna Nuclear Power Plant. ML17309A4481990-08-16016 August 1990 Responds to NRC 900717 Ltr Re Violations Noted in Insp Rept 50-244/90-80.Corrective Actions:Westinghouse Drawing E-2508 Approved & Issued to Central Records ML17261B1351990-08-15015 August 1990 Provides Info & Assessment on Integrity of Connection of Containment Bldg to Foundation,Per 900808 Telcon.Util Believes That Existing Condition Not Safety Concern & No Reason Exists to Suspect Joint Will Not Perform Function ML17261B1361990-08-14014 August 1990 Responds to Commitment Tracking Concerns Noted in Insp Rept 50-244/90-09 & Planned Corrective Actions.Util Confirms Commitments Dates for Implementation of Effective Shelf Life Program & Comprehensive Preventive Maint Program for Parts ML17261B1331990-08-13013 August 1990 Recommits to Performing Enhanced primary-to-secondary Leak Rate Monitoring,Per NRC Bulletin 88-002 ML17261B1261990-07-30030 July 1990 Clarifies Commitment Made in 900316 RO Re Restoring Inoperable Fire Damper I-411-21-P.Util Plans to Design Removable Track Which Will Allow Charcoal Drawer to Be Manipulated.Definitive Schedule Will Be Provided in 60 Days ML20055H9291990-07-23023 July 1990 Forwards Revised Page 6:4 of Plant Security Plan.Page Withheld Per 10CFR73.21 ML17261B1091990-07-20020 July 1990 Advises That Structural Evaluations of Containment Sys Being Performed in Response to Containment Integrity Insp ML17309A4471990-07-17017 July 1990 Forwards Decommissioning Rept, for Plant,Per 10CFR50.33(k) & 50.75(b) ML17261B0881990-07-13013 July 1990 Provides Update to Util 860616 Ltr Re Implementation of NUREG-0737,Item 6.2,Suppl 1, Emergency Response Capability. ML17261B0921990-07-11011 July 1990 Responds to 900709 Request for Addl Info Re Containment Integrity Insp.Util Will Provide Preliminary Results by 900716 Re Where Groundwater Entering Annular Access Area. Meeting Proposed for Wk of 900723 ML20044B0481990-07-10010 July 1990 Discusses Review of Station Blackout Documentation,Per 10CFR50.63.Licensee Will Complete Enhancements to Station Blackout Documentation Identified in Attachment 1 as Indicated & Other Items Will Be Completed within 2 Yrs ML17309A4461990-07-0909 July 1990 Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues Resolved W/ Imposition of Requirements or Corrective Actions ML17261B0931990-07-0909 July 1990 Responds to Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment. Util Completed All Actions Requested in Generic Ltr & Will Retain Documentation Verification for Min of 2 Yrs ML17261B0901990-07-0909 July 1990 Advises That Licensee Will Install Containment Isolation Signal Going to Valve AOV 745 by End of 1992 Refueling Outage,Per Util 900608 Ltr Notifying of Condition Outside Design Basis of Plant Under 10CFR50.72 Criterion ML17250B2151990-06-29029 June 1990 Forwards Application for Amend to License DPR-18, Reformatting Auxiliary Electrical Sys Tech Specs & Action Statements for Offsite & Onsite Power Sources Available for Plant Auxiliaries ML17261B0851990-06-28028 June 1990 Forwards LER 89-016-02 Re 891117 Failure of Safety Injection Block/Unblock Switch Which Could Render Both Trains of Safety Injection Sys Inoperable.Also Reported Per Part 21 ML17250B1991990-06-0808 June 1990 Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983 ML17250B1951990-06-0505 June 1990 Responds to Generic Ltr 89-08, Erosion/Corrosion Induced Pipe Wall Thinning. Util Developed Erosion/Corrosion Program for Single & Two Phase Sys Consistent W/Requirements of NUREG-1344 & NUMARC 870611 Rept ML17250B1851990-06-0101 June 1990 Forwards Application for Amend to License DPR-18,providing Guidance for Action Statements Associated W/Power Distribution Limit Specs ML17261B0691990-06-0101 June 1990 Discusses Testing Frequency for Insp of Incore Neutron Monitoring Sys Thimble Tubes,Per NRC Bulletin 88-009.Thimble Tube Indicating Greatest Wear Recently Repositioned in Effort to Minimize Future Wear ML17250B1811990-05-31031 May 1990 Forwards Addl Info Re Response to Notice of Violation from Insp Rept 50-244/90-04.Info Withheld (Ref 10CFR73.21) ML17250B1801990-05-29029 May 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Util Does Not Need to Develop Enhanced Surveillance Program to Monitor Currently Installed Transmitter Based Upon Limited Installed Quantity ML17250B1841990-05-29029 May 1990 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel (SS) Internal Pre-Load Bolting.... Valves Disassembled & 410 SS Studs Removed & Visually & Liquid Penetrant Examined ML17261B0961990-05-29029 May 1990 Responds to NRC 900427 Ltr Re Violations Noted in Insp Rept 50-244/90-07.Corrective Actions:Worker Records Immediately Corrected Indicating That Worker Received No Significant Exposure for Time While Error Occurred ML17250B1721990-05-18018 May 1990 Forwards Rev 1 to Summary Exam Rept for 1990 Steam Generator Eddy Current Insp. Rept Summarizes Observations & Corrective Actions Resulting from Insps Performed During 1990 Outage ML20042G8931990-05-0808 May 1990 Forwards Revised Emergency Operating Procedures,Including Rev 3 to AP-RHR.2,Rev 1 to ES-0.3 & Rev 11 to ES-1.3 ML17261B0611990-04-26026 April 1990 Advises of Changes to 890201 Commitments Made Under Programmed Enhancement Response to Generic Ltr 88-17.Changes Will Not Reduce Capability to Operate Safely in Reduced Inventory Condition or Scope of Programmed Enhancements ML17261B0591990-04-17017 April 1990 Advises That No Interlocks Required for RHR motor-operated Valves 701 & 720 Based on Present Arrangement.Listed Failures Would Have to Occur in Order for Potential Overpressurization of RHR Sys to Occur ML17261B0461990-04-12012 April 1990 Responds to Issues Discussed During 900323 Telcon Re Inservice Testing Program Status & Relief Request.Current Test Methodology for Seat Leakage Acceptable Based on Application of Direct Measurement Sys ML17262A1431990-04-12012 April 1990 Responds to NRC 900202 Ltr Re Weaknesses Noted in Insp Rept 50-244/89-80.Corrective Actions:Procedure ES-0.3 Modified to Provide Guidance for Rapid Cooldown & Depressurization W/ & W/O Reactor & Vessel Instrumentation Sys ML17250B1451990-04-0606 April 1990 Discusses Impact of SER Issuance for Inservice Insp & Inservice Testing Programs & Advises That Timing Will Not Affect Util Implementation Plans for Programs,Except as Listed ML17250B1441990-04-0505 April 1990 Responds to NRC 900309 Ltr Re Violations Noted in Insp Rept 50-244/90-02.Corrective Actions:Positive Indicator Will Be Installed for All Check Valves to Enhance Visible Positive Position Verification Ability & to Avoid Confusion ML17261B0401990-03-30030 March 1990 Advises That Final Results of Station Blackout Documentation Review Will Be Submitted to NRC on or About 900501 ML17261B0321990-03-28028 March 1990 Forwards Annual Rept of ECCS Model Revs as Applicable to Facility,Per 10CFR50.46.Mods to Model for Small Break LOCA Do Not Affect Calculated Peak Clad Temp ML17261B0331990-03-28028 March 1990 Forwards Info Re Reactor Vessel Issues,Per 900305 Telcon Concerning Change of License Expiration Date ML17261B0231990-03-26026 March 1990 Responds to NRC 890222 Ltr Re Violations Noted in Insp Rept 50-244/89-17.Corrective Actions:Personnel Verified Safety Injection Block/Unblock Switch in Proper Position & Operator Procedure 0-1.1 Changed as Indicated ML17261B0281990-03-23023 March 1990 Forwards Addl Info Re Proposed Tech Spec Amend Concerning Use of Reconstituted Fuel,Per 900315 Telcon.During Fuel Assembly Reconstitution,Failed Fuel Rods Will Be Placed W/ Filler Rods ML17261B0221990-03-22022 March 1990 Provides Revised Test Schedule for motor-operated Valve Diagnostic Test Program,Per IE Bulletin 85-003.NRC Notification of Changes to Valve Operability Program Required by Generic Ltr 89-10,dtd 890828 ML17261B0411990-03-20020 March 1990 Forwards Summary of Onsite Property Damage Coverage Currently in Force at Plant,Per 10CFR50.54(w)(4) ML17261B0151990-03-19019 March 1990 Responds to Generic Ltr 89-19, Safety Implication of Control Sys in LWR Nuclear Plants (USI A-47). Overfill Protection Provided Through Trip Bistables in Reactor Protection Racks Powered from 120-volt Instrument Buses ML17261B0101990-03-15015 March 1990 Forwards Application for Amend to License DPR-18,allowing Use of Reconstituted Fuel Assemblies ML20012D2831990-03-14014 March 1990 Responds to Issues Discussed During 900307 Telcon W/Util & Eg&G Re Inservice Testing Program Status & Relief Request. Valves 5960A & 5960B Will Be Disassembled to Verify Forward Flow.Relief Requests PR-8 & VR-25 Encl 1990-09-18
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ACCELERATED DISTRIBUTION DEMONST$&TION SYSTEM REGULATORY INFORMATION DISTRXBUTION SYSTEM (RIDS)
ESSION NBR:9004040007 DOC ~ DATE: 90/03/26 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp.
RECIP.NAME RECIPIENT, AFFILIATION RUSSELL,W.T; Region 1, Ofc of the Director R
SUBJECT:
Responds 50-244/89-17.
to NRC 890222 ltr re violations noted in Insp Rept DISTRXBUTION CODE: IE01D COPIES RECEIVED:LTR ENCL 0 SIZE:
TITLE: General (50 Dkt)-Insp Rept/Notice of Vi lation Response, NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).. 05000244,']
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 PD JOHNSON,A INTERNAL'EOD 1 AEOD/DEIIB 1 AEOD/TPAD 1 DEDRO 1 NRR SHANKMAN,S 1 NRR/DLPQ/LPEB10 1 NRR/DOEA DIR 11 1 NRR/DREP/PEPB9D 1 NRR/DREP/PRPB11 ,2
'1 NRR/DRIS/DIR 1 NRR/DST/DXR 8E2 NRR/PMAS/ILRB12 l NUDOCS=ABSTRACZ 1 OGC/HDS2 1 REG FIXE'- ~02~ 1 RES MORISSEAU,D 1 RGN1 FILE 01 1 EXTERNAL: LPDR 1 NRC PDR NSIC 1 legs p]5 7~ '-'
.A NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WAS'ONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
OTAL NUMBER OF COPIES REQUIRED: LTTR 23 ENCL
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ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER, N.Y. 14849-pppg March 26, 1990 TCKCRHONC ARCA COOK 71K 546 2700 Mr. William T. Russell Regional Administrator U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, Pennsylvania 19406
Subject:
Response to Notices of Violation Inspection Report No. 50-244/89-17 R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Russell:
This letter is in response to the February 22, 1989 letter from Jon R. Johnson, Chief, Projects Branch No. 3 to Robert E. Smith, Senior Vice President, RG&E, which transmitted Inspection Report No. 50-244/89-17. In that report, two violations were identified. The following provides a reply to the violations pursuant to 10 CFR 2.201.
RESTATEMENT OF VIOLATIONS During inspection at the R.E. Ginna Nuclear Power Plant from December 12, 1989 through January 8, 1990, the following violations were identified and evaluated in accordance with the NRC Enforcement Policy (10 CFR 2, Appendix C):
A. 10 CFR 50, Appendix B, Criterion XVI, and the Ginna Quality Assurance Manual, Section 16, require prompt identification and correction of conditions adverse to quality including failures, malfunctions, deficiencies, defective material and equipment, and nonconformances.
Contrary to the above, a safety injection system design deficiency was not promptly identified and corrected when corporate engineering was notified on or before October 20, '989 that failure of the safety injection block/unblock switch could block automatic safety injection actuation on low pressurizer pressure or low steam line pressure. Corporate engineering did. not conclude that this problem existed at Ginna until about November 17, 1989, and site technical personnel were not informed about the deficiency until December 19, 1989.
This is a Severity Level IV violation (Supplement I).
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4 B. 10 CFR 50, Appendix B, Criterion V, and the Ginna Quality Assurance Manual, Section 5, require activities affecting quality to be accomplished in accordance with instructions, procedures, or drawings which include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
Contrary to the above, on December 15, 1989, maintenance was performed on a safety-related motor-operated valve in the safety injection system in accordance with a procedure which included an inappropriate torque specification.
This is a Severity Level V violation (Supplement I).
RESPONSE TO VIOLATION A RG&E Position on Existence of Violation Rochester Gas and Electric Corporation (RG&E) concurs that a violation of Appendix B, Criterion XVI occurred. RG&E recognizes that communication between corporate engineering and site personnel on issues of potential safety significance should be formalized. Our efforts to address this concern are provided in Section 4, "Long Term Enhancements". As explained below, RG&E also believes that with respect to the issue identified on October 20, 1989, we acted in a manner consistent with the safety
.significance of the matter.
- 2. Reason for Violation As Inspection Report No. 50-244/89-17 (p. 7) indicates, RG&E received notice on October 20, 1989, from Westinghouse Electric Corporation (Westinghouse) of an apparent generic design deficiency related to the type of safety injection (SI) block/unblock switch used at various Westinghouse reactors. The Westinghouse letter, dated October 12, 1989, concluded that a "single failure of the switch (Westinghouse OT2) could block either the automatic low pressurizer pressure or the low steamline pressure SI signal in both trains" [emphasis supplied]. The letter also' stated that the probability of switch failure was "10 10 '/yr" :and that, while a design change was recommended, the situation was "not an immediate safety concern."
In addition, the Westinghouse letter referred to a Licensee Event Report (LER), No. 88-007-00, submitted by Wisconsin, Electric Power Company (Wisconsin Electric) on September 16, 1988, concerning the same issue at the Point Beach Nuclear Plant (Point Beach). The Wisconsin Electric LER concluded that "this condition will not have a significant impact on the health and safety of the general public or the employees of the Point Beach Nuclear Plant."
The LER noted that the Point Beach facility was operating at 100% capacity when the concern was identified and that design change would not'e made until the next scheduled outage.
Upon receipt of the Westinghouse notification on October 20, 1989, RG&E (corporate) initiated a timely review for applicability to Ginna Station. Based on the Wisconsin Electric LER and on Westinghouse's calculation of the low probability of switch failure, it was apparent that the matter did not constitute an immediate safety concern.
When it was identified that the switch configuration was applicable to Ginna Station, an internal engineering recommendation was made consistent with the guidance of the Westinghouse letter and attached LER, that an EWR be initiated. This was completed on November 17, 1989. This recommendation was then evaluated within Nuclear Safety and Licensing, resulting in a discussion with site technical support personnel relative to this situation on December 19, 1989. On December 20, site personnel initiated a Ginna Station Event Report per Procedure A-25.1 (Event No.89-168). The event report indicated that the site Plant Operations Review Committee (PORC) had, on December 20, 1989, concluded that plant operation could continue for the following reasons:
very low (i.e., 10 'o
- 1. Westinghouse stated that the. probability of failure was 10 '/yr);
- 2. Emergency Operating Procedures directed Operators to use manual SI initiation where indicators show automatic initiation has failed;
- 3. A separate automatic SI initiating mechanism would activate when containment pressure reached 4 psig;
- 4. During depressurization, a bistable light will'lert operators of a blocked SI signal; and
- 5. Visual verification of the SI switch plunger position indicates that the contacts are in the proper position.
The violation states that the time between October 20, 1989, when RG&E (corporate) was notified by Westinghouse, and the communication of this information to the site technical staff on December 19, 1989, shows that the SI design deficiency was not promptly identified and corrected, and indicates problems in communication between corporate engineering and site personnel. While RG&E does not deny this violation, we believe that the actions taken by RG&E were appropriate in view of RG&E's preliminary conclusion that the issue did not constitute an immediate safety concern.
RG&E believes that Appendix B, Criterion XVI does not establish a precise time limit for resolution of safety issues. Rather, issues such as "promptness" or "timeliness" are subjective matters that inherently depend upon the safety significance of the situation. Given that RGGE had a documented recommendation from Westinghouse that no immediate safety concern existed (as corroborated by the Point Beach LER), its actions toward resolution of the issue were prompt and timely. Any other interpretations of Criterion XVI would be counter to public health and safety because it would require licensees to treat all deficiencies or non-conforming items the same (i.e., regardless of safety significance).
This same basic philosophy was affirmed in an analogous context 'in recent guidance issued by NRC's Office of Nuclear Reactor Regulation '(NRR). Specifically, on July 19, 1989, Dr. T.E. Murley, Director, NRC/NRR, sent a memorandum to all of the regional administrators entitled "Guidance on Action To Be Taken Following Discovery of Potentially Nonconforming Equipment." In his memorandum, Dr. Murley stated that "[t]here is no generally appropriate timeframe in which operability determinations should be made." For equipment which is "clearly inoperable," an immediate declaration of inoperability should be made and the appropriate technical specifications followed. However, Dr. Murley's memorandum contrasts this situation with those where equipment nonconformances simply raise the issue of operability.
In such situations Dr. Murley states that:
operability determinations should be made by licensees as soon as racticable, and in a timeframe commensurate with the a licable e ui ment's im ortance to safet usin the best information available,(e.g., analyses, a test or partial test, experience with operating events, engineering judgement or a combination of the factors) (emphasis supplied).
Although this guidance relates to timing of operability determinations, it is equally appropriate with respect to resolution of open items under Criterion XVI. Consistent with this philosophy and based on the best information available, future cases of this type will be resolved "as soon as practicable" and in a time commensurate with the safety significance of the matter. Communication between corporate and site personnel will be initiated promptly once applicability to Ginna Station is determined.
Corrective Ste s Which Have Been Taken and the Results Achieved Corporate and site technical staff and the PORC have reviewed the circumstances surrounding the potentially generic design deficiency related to the control room SI block/unblock switch. As stated in LER 89-016, the.
following actions were taken:
Knowledgeable personnel inspected the plunger position of the SI Block/Unblock Switch and verified that the switch contacts were in the proper position.
~ Operating Procedure 0-1.1 (Plant Heatup From Cold Shutdown to Hot Shutdown) was changed to add the following note and check-off to Step 5.11.6:
NOTE: Prior to placing the SI Block/Unblock Switch to the normal position, station an operator inside the MCB in direct observation of the SI Block/Unblock Switch to observe that both plunger tips are recessed inward after the switch is placed. to normal position.-
Block switch plunger t'ips position inward
~ An RG&E operator aid tag was placed on the .MCB adjacent to the SI Block/Unblock Switch denoting the note- from 0-1.1.
~ An RG&E operator aid tag was also placed inside the MCB adj acent to the rear of the SI Block/Unblock Switch stating the following: This is the switch we verify that the plunger's tips are recessed inward when the switch is placed to normal (labeled LAK) .
A spare switch of similar design has been placed in the Control Room for the purpose of training the operators to recognize the differences in plunger position.
These actions are considered adequate to provide reasonable assurance of SI system operability until the situation can be permanently dispositioned. Finally, EWR 5025 was initiated to provide for the installation of independent SI block/unblock switches for each SI train which is planned for the 1991 refueling outage.
- 4. Corrective Ste s Which Will Be Taken to Avoid Further Violation RG&E has recently taken steps to upgrade the overall corrective action program for Ginna Station. The need for improvements was noted during the course of the RHR System Safety System Functional Inspection (SSFI), and is also considered appropriate due to RG&E's initiation of a comprehensive Configuration Management/Design Basis Program. We are working with the NUMARC Design Basis Issues Working Group to develop an improved problem identification and resolution program.
The improved program will:
~ Improve the process of identifying, analyzing, and resolving problems;
~ Improve the RG&E internal review process, including formalized means of communication between corporate engineering and site personnel on issues of potential safety significance; and Part of the implementation of this effort will include specific procedural upgrades, enhancement of our corrective action tracking system, and the issuance of a corporate policy which addresses problem identification and reporting. We believe that this broad effort, when fully implemented, will improve our capability to consistently identify and disposition potential safety issues commensurate with their significance.
- 5. Date When Full Com liance Will Be Achieved Long term and short term actions and schedules have been described above. Formal guidance concerning communication between corporate and site personnel on identified problem issues is under development, and is targeted for completion by July 1990.
RESPONSE TO VIOLATION B Rochester Gas and Electric concurs with this violation as stated below.
Reason for Violation Rochester Gas and Electric agrees that, Ginna Station does not have an established written policy regarding consideration of inherent inaccuracy of calibrated measuring and test, equipment (M&TE) when developing acceptance criteria.
As- a common practice, torquing methods address only instru-ment "indication" and are not meant to include the instrument accuracy. This practice is based on the fact that torque is only a general indicator of bolting pre-load because of the inaccuracies, e.g., lubrication, thread fit, thread condition, etc., inherent in the torque equation. When highly accurate bolt pre-loading is required, means other than torque is used, i.e., stud elongation to determine bolt pre-load.
The Corrective Ste s Which Have Been Taken and the Results Achieved Due to the successful completion of post maintenance testing, no action regarding the valve packing adjustment has been taken.
A-1603.4, "Work Order Scheduling" was revised to require work and testing to be completed on individual trains prior to starting maintenance on a redundant train.
' The Corrective Ste s Which Will Be Taken to Avoid Further Violation
- 1. Administrative procedure A-1603.3, "Work Order Planning" will be revised to state a Ginna Station policy regarding consideration of M&TE inherent inaccuracy and provide direction for development'f acceptance criteria utilizi'ng this equipment.
- 2. A new procedure for packing adjustment is being developed to provide specific direction for adjustment of valves repacked under the Valve Packing Improvement Program and to provide a method of maintaining and updating valve packing data.
The Date When Full Com liance Will Be Achieved The anticipated effective date of the above procedures is May 1, 1990, for the maintenance procedures and June 30, 1990, for the administrative procedure.
Very truly yours, Robert C. Me dy Division Manager Nuclear Production GJWN093 Enclosures xc: U.S. Nuclear Regulatory Commission (original)
Document Control Desk Washington, D.C. 20555 Allen R. Johnson .(Mail Stop 14D1)
Project Directorate I-3 Washington, D.C. 20555 Nicholas S. Reynolds, Esq.
Bishop, Cook, Purcell and Reynolds 1400 L. Street, N.W.
Washington, D.C. 20005-3502 Ginna NRC Senior Resident Inspector
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