ML20199F666
| ML20199F666 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 11/17/1997 |
| From: | Tulon T COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-456-97-12, 50-457-97-12, NUDOCS 9711240296 | |
| Download: ML20199F666 (17) | |
Text
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November 17,1997 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Reply to Notice of Violation NRC Inspection Report 50-456(457)/97012 Braidwood Nuclear Power Station Units 1 and 2 NRC Docket Numbers 50-456 and 50-457
Reference:
J. A. Grobe letter to T. J. Tulon dated October 17,1997, transmitting Notice of Violation from inspection Repuit 50 456(457)/97012 During a three week inspection penod which ended on August 15, 1997, an evaluation of Braidwood Station's Engineering and Technical Support area was conducted. During the inspection, six Severity Level IV violations were identified. These violations are documented in the reference specified above. Comed's response to the violations is included in the attachment to this letter. Braidwood Station concurs with the cited violations, with the exception of one case. In Violation 2b, the station believes that a safety evaluation was not required for a revision to Braidwood's Onsite Review procedure. The decision to not perform the evaluation is consistent with available regulatory and industry guidance.
Overall, Braidwood Engineering has exhibited sustained performance improvements, and efforts continue to address identified weaknesses. Engineering personnel have provided good support to plant activities as evidenced by reductions in Operator Workarounds, Temporary Modifications and the Engineering Request backlog, and aggressive actions are being taken to resolve concerns associated with attention to detail. Specific surveillances are being reviewed and revised as y necessary to enhance their clarity.and ease of execution. In addition, training is also being Q-enhanced to ensure personnel are properly qualified for their assigned tasks. *These efforts are i expected to improve perfomiance. 3
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9711240296 971117 i
PDR ADOCK 05000456 O PDR A tinkom Compan)
Document Control Desk November 17,1997 Page 2 The following commitments were made in the attachment to this letter:
- NRC Information Notice IN 9716, " Preconditioning of Plant Structures, Systems, and Components Before ASME Code Inservice Testing or Technical Specification Surveillance Testing," was routed to Maintenance, Engineering, and Operations personnel to review for applicability. These departments will review procedures in their functional areas for potential preconditioning concerns. When the review is completed, procedux concerns will be resolc appropriately.
- A tailgate discussion will be held with site personnel who may be involved with Technical Specification surveillances or activities which may affect the results of Technical Specification surveillances from the preconditioning aspect. The discussion will include guidance for recognizing and reporting preconditioning issues.
If ycur staff has any questions or comments concerning this letter, or would like to further discuss the details included in this response, please contact Terrence Simpkin, Braidwood Regulatory Assurance Supervisor, at (815) 458-2801, extension 2980.
M ,h Timothy J. Tulon Site Vice President Braidwood Nuclear Generating Station Attachment ec: A.B. Beach. NRC Regional Administrator, Region 111 G.F. Dick, Jr., Project Manager, NRR C.J. Phillips, Senior Resident Inspector F. Niziotek, Division of Engineering, Office of Nuclear Safety, IDNS ohrc97125m doc
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, ATI ACl! MENT I REl'LY TO NOTICE OF VIOt,ATION VIOLATjON f $0-456(457F97012-01)
- 1. 10 CFR 50, Appendix II, Criterion XI, " Test Control," requires that a test program be-established to assure that all testing required to clemonstrate that structures, systems, and components (SSCs) will perform satisfactorily in service is identified and perfonned in accordance with written test procedures which include provisions for assuring that the testing is performed under suitable environmental conditions.
Contrary to the above:
- a. The licensee failed to perform Braidwood Electrical Maintenance Department Surveillance (BwilS) 4009 035, " Containment Penetration Conductor !
Overcunent Protective Devices From 480 Volt Switchgear," for the 2D reactor containment fim cooler (RCFC) high speed fan breaker and the 1D RCFC low speed fan breaker on February 28, 1996 and April 19, 1997, under suitable ;
environmental conditions. Specifically, the breakers were manually cycled to ensure no excessive binding or friction existed in the operating mechanism just prior to testing the overcurrent protective devices. As a result, the environment was altered and the data obtained did not represent testing under suitable environmental conditions.
REASON FOR Tile VIO12 TION Dw!!S 4009-035 " Containment Penetration Conductor Overcurrent Protective Devices from 480 Volt Switchgear" Procedure, is used to check the operability of all containment conductor overcurrent protective devices from 480 volt switchgear on a sampling basis. This procedure required the 480 Volt breaker to be cycled prior to testing the overcurrent device. The action to cycle the breaker prior to testhg the overcurrent device was based on the manufacturer's recommendation to cycle the breaker as a method of redistributing the lubricants and increase the reliability of the breaker, llraidwood Station concurs that this action potentially preconditioned the breaker prior to testing the overcurrent device. The preconditioning effects were not recognized at the time the procedure was generated.
CORRECTIVE ACTIONS TAKEN AND RESULTS ACil'EVED A review of"as-lcll" values for the surveillance was performed. The data was found to satisfy the specified acceptance criteria.
IlwliS 4009-035 was revised to eliminate cycling the breaker prior to testing the overcurrent device, i
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ATTACilMENT I REPLY TO NOTICE OF VIOLATION VIOLATION (50-456(457P97012-01)
DATE WilEN FULL COMPLIANCE WAS ACillEVED Full compliance was achieved when the data obtained in the surveillance was evaluated and determined to be acceptable.-
VIOLATION (50-456(457%97012-01b) l b. The licensee failed to assure that degraded voltage relay testing for divisions 11, 12, 21, and 22 was performed under suitable environmental conditions.
Specifically on April 6,1997; April 20,1997; April 8,1996; and March 26,1997; division 11,12, 21 and 22 degraded voltage ti'.ners were preconditioned by performing BwHS 4002-091, "T'me Delay Pslay Surveillance," just prior to performing Braidwood Surveillance Procedure (BwVS) 3.2.2-4, "l?nderk.tage Time Response 18-Month Surveillance," to meet requirements set forth in Technical Specification Table 4.3 2.
REASON FOR TIIE VIOLATION in 1992, a concem was identified at other Comed Stations regarding the method used to verify setpoints on safety related time delay relays. As a result of the concern, Braidwocd Station developed piocedural guidance and performed a setpoint verification of time delay relays. The purpose of the test was to gather baseline data to determine the drift characteristics exhibited by the relays over time. This was done in order to ascertain the appropriate preventive maintenance frequency. The preconditioning resulted from the manner in which the performance of surveillances BwilS 4002-091 ar.d BwVS 3.2.2-4 were sequenced. Because BwllS 4002-091 was performedjust prior to BwVS 3.2,2-4, the as found data of the BwVS was preconditioned.
CORRECTIVE ACTIONS TAKEN AND RESULTS ACillEVED
- A review of"as-left" values for the surveillance was performed. The data w. found to satisfy the specified acceptance criteria.
The testing of the undervoltage time delay relays was removed from the Electrical Maintenance Surveillance (BwilS 4002-091, " Time Delay Relay Surveillance"). This testing is incorporated
-in BwVS 3.2.2-4 which was also revised to include steps to record the "as-found" and "as left" response time values.
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,- ATTACilMENT 1 11EPLY TO NOTICE OF VIOLATION -
' VIOLATION (50456f 457P97012-01) - .
DATE WIIEN FULL COMPLIANCE WAS AClllEVED Full-compliance was achieved when the data obtained in the surveillance was evaluated and determined to be acceptable.
ACTIONS TO IlE TAKEN TO PREVENT RECURRENCE (1a and 1b)
NRC Information Notice IN 9716, " Preconditioning of Plant Structures, Systems, and Components llefore ASME Code inservice Testing or Technical Specification Surveillance Testing " was routed to Maintenance, Engineering, and Operations personnel to review for applicability. These departments will review procedures in their functional areas fbr potential precond;tioning concerns. When the review is completed, procedure concerns will be resolved appropriately, A tailgate discussion will be held with site personnel who may be involved with Technicel Specification surveillances or activities which may affect the results of Technical Specification surveillances from the preconditioning aspect. The discussion will include guidance for recognizing and reporting preconditioning issues.
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, ATTACllMENT I -
REPLY TO NOTICE OF VIOLATION :
VIOLATION (50-454 tS7P07012-02) 2 . 10 CFR 50.59(a)(1) requires, in partithat a licensee may make changes in the facility as -
- described in the safety analysis report, make changes in the procedures as described in the -
safety analytis report; and conduct- tests _ or- experimen'.3 not described in the_ safety _ _
- analysis report, without written Commission approval, unless the proposed change, testi or experiment ' involves . a : change in the licensee's technical specifications or an unreviewed safety question.
10 Cl R 50.59(b)(1) requires, in part, that the licensee shall mahnain records of changes in' the facility and of changes in procedures made pursuant to this section and that these ,
records must include a written safety evaluation which must provide the basis for a determination that the change, test, or experiment does not involve an unreviewed safety .l question.
- n. Braidwood Updated Final Safety Analysis Report (UFSAR), Revi: ion 6, Section 6.3, " Emergency Core Cooling System," states that relief valves are installed in various sections of the emergency core cooling system (ECCS) to protect lines which have a lower design pressure than the reactor coolant system. In addition, UFSAR Section 6.3, Table 6.3-2, "ECCS Relief Valves Data," lists the ECCS system relief valver with their capacities and setpoints. Table 6.3-2 includes the :
safety injection (SI) pump discharge relief valves.
Contrary to the above:
On July 30,1996, a 10 CFR 50.59 safety evaluation to gag closed safety injection pump discharge relief valve 2S18851 failed to provide an adequate basis, in writing, that the change did not involve an unreviewed safety question.
Specifically, the evaluation failed to address the fact that American Society of Mechanical Engineers (ASME) Section ill Code require.nents for relief valves (Article NC-7000) would not be met with 2S18851 gagged closed.
REASON FOR Tile VIOL ATION
'An evaluation pursuantf to 10 CFR 50.59 had previously been written to document the nonconfonning. condition of the 2S18851 (gagged- closed). Although the ASME Code requirements were recognized and discussed by Station personnel at the time the 50.59 was written, the discussion of the ASME Section III Code requirements for the relief valve was not documented in_ the. 50.59.- In retrospect, the use of a-50.59 evaluation for the purpose of documenting the nonconforming condition of the 2SI8851 was not appropriate. Consistent with the guidance of GL- 91-18, an operability determination had been performed which documented the condition and its effect on operability. Because the intent was to restore the valve to its
. previous condition as described in the UFSAR at the next available opportunity (no later than the 4
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.. ATTACilMENT l-
- REPLY TO NOTICE OF VIOLATION QLAllON (50-45.6(457197012-02)
.next refueling outage), a 50.59 evaluation was not required. This guidance has been reaffirmed in NEl 96-07 (Final Draf1) and GL 91 18, Revision 1. ;
CORRECTIVE ACTIONS TAKEI4 AND RESULTS ACillEVED The failure to reference the ASME Code requirement in the 50.59 was discussed with the 50.59 preparer t.nd reviewer. In addition, discussions on this subject were held with Site Engineering personnel. The discussions emphasized the need to completely document the rationale for the acceptability of the change being evaluated when performing a 50.59 evaluation.
The 2S18851 relief valve was .5placed and satisfactorily tested during the A2R06 refueling outage.
ACTIONS TAKEN TO PREVENT RECURRENCE The need for thorough and complete documentation in 50.59 evaluations was re-emphasized to all engineers as part of training asociated with the implementation of procedure NSWP A-04, "10 CFR 50.59 Safety Evaluation Process," in January 1997 and again as part of third quarter 1997 continuing training for the Engineering Support population.
The Braidwood procedure BWAP 330-10," Operability Determinations," was revised to provide additional guidance on the differences between an operability determination and a 50.59 evaluation and the situations in which they should be used. This was completed in May 1997.
Training on this point was provided to all engineers initially as part of a Regulatory Fundamentals course in the thiid quarter of 1996, and again as part of training associated with the operability determination procedure change in the third quarter of 1997.
DATE WilEN FUI L COhlPI IANCE WAS'ACIIIEVFD Full compliance was achieved when the relief valve was replaced. This replacement corrected the situation evaluated by the inadequately documented safety evaluation by returning the station to the configuration addressed in the Updated Final Safety Analysis Report (UFSAR).
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[ A1TACliMENT 1 REPLY TO NOTICE OF VIOLATION -
VIOLATION (50-456(457W7012-02)
VIOL ATION (50-456(457)/97012-02b) 2b. 10 CFR 50.34(b)(6)(ii), " Contents of Applications; Technical Information," states that the Final Safety Analysis Report shall include managerial and administrative controls to be used to assure safe operation and that the information on the
- controls to be uced for a ncclear power plant shall include a discussion of how the applicable requirenien'.s of 10 CFR 50, Appendix B," Quality Assurance Criteria for Nuclear Power Plants and Fuel Proce sing Plants," will be satisfied.
Braidwood UFSAR, Revision 6 Chapter 17, " Quality Assurance," states that the -
Braidwood quality assurance program is conducted in accordance with the Commonwealth Edison Quality Assurance (QA) Program for Nuclear Generating Stations which was submitted to and approved by the NRC as Topical Report CE- 1.
Braidwood UFSAR, Revision 6, Chapter 17, " Quality Assurance," also states that Commonwealth Edison Topical Report CE-l-A is the basis for the QA program at braidwood.
Commonwealth Edison Topical Report CE-1 A, Revision 65f, Seetion 20, Paragraph 3.31, states that the station manager shall independently review and aporove the findings and recommendations developed by personnel performing the Onsite Review and Investigative (OSR) Function.
Contrary to the above:
On March 31, 1997, the licensee failed to perform a 10 CFR 50.59 safety evaluation to revise Braidwood Administrative Procedure fBwAP) 1205-3, "Onsite Review and Investigative Function," to allow the Plant Onsite Review Committee (PORC) to fulfill the OSR function although the station manager was the chaimian of the PORC and his/her ability to independently review and approve the findings and recommendations developed by personnel performing the OSR function as required by Topical Report CE-1-A could be affected.
REASON FOR Tile VIOLATION Braidwood Station does not believe that the cited violation represents a violation of 10 CFR 50.59. In March,1997, BwAP 1205-3, "Onsite Review and investigative Function" Procedure, was revised. One of the changes allowed the PORC Committee to fultill the Onsite Review and investigative function when formally specMied. A 50.59 screedng had been performed to evaluate this procedure change which indicated that a full evaluation was not required. This was 6
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,' AlTACilMENT I I
RI.I'LY TO NOTICI Of VIOLATION y10LAllON f 30-456(457r97012-02 n consistent with industry guidance (NSAC-125, Section 4.1.2) which indicates that changes to procederes not outlined, summarized, or completely described in the UFSAR do not require safety evaluations.
Comed's Safety Evaluation Procedure, NSWP A 04, states that Quality Assurance program descript!ons and the Quality Assurance Topical Report (QATR) are information which are "part of the licensing basis and may be a part of the UFSAR; however chenges are governed by other regulations. Cen.pliance with the other specifically applicable regulations constitutes ccmpliance with 50.59." The reference used as a basis fbr this statement was NRC's Policy Statement on Tect nical Specification improvement (Federal Register 39134 dated July 22, 1993). Additional guidance is provided in NUREO 1606, " Proposed Regulatory Guidance Related to implementation of 10 CFR 50.59 (Changes, Tests, or Experiments)". This document states, " Specific Ilcensee programs, such as emergency preparedness plans, security plans, and quality assurance plans have change control processes explicitly established by regulation (in 10 CFR 50.54) even though the plans may also be referenced by the SAR. These specific change control processes are considered applicable to the plans rather than the 10 CFR 50.59 process because the 10 CFR 50.54 processes generally contain more restrictive reporting requirements and different reporting criteria for determining when prior staff approval is needed."
Finally, NEl 96 07, the industry's draft guidance document, reaffirms the position specified in NSAC 125, Figure 1.1. This document states, "A licensee's Quality Assurance Manual, Emergency Plan, and Security Plan, are controlled by separate NRC requirements (e.g. 50.54 and 50.55a). A 50.59 review does not need to be conducted on changes to these documents as long is other parts of the SAR are not changed or impacted."
CORRECTIVE ACTIONS TAKEN AND RIISULTS AClllEVED lhaidwood Station has reviewed its Onsite Review (OSR) and PORC processes and perfonnance and has not identified an occurrence where the ability of the Station Manager to maintain the oliectivity necessary to independently review and approve OSR findings has been compromised.
A review of guidance documents was conducted to ensure proper compliance.13ased on this review,ik.ildwood Station has concluded that a 10 CFR 50.59 safety evaluation was not required fbr the revision to the station's Onsite Review procedure.
ACTIONS TAKEN TO PREVENT RECURRENC E liraidwood provided comments on NUREG-1606 (which was issued for comments) and will continue to support the industry's efforts to achieve consensus with the NRC on a common 50.59 guidance document.
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- I NITACliMI;NT I Riil'IN 10 NOllCE Of VIOLAllON ,
VIOLNIlON (50 456(457)97012-02) j DATl! Will!N FUl_L COMPI_lANCII WAS AGilEVED ,
Following a review ofindustry guidance documents, including NSAC 125, NURl!G 1606, and -
Niil 96 07, liraidwood Station has concluded that compliance was properly maintained by perfonning a 10 CFR 50.59 screening to evaluate a revision to the Station's Onsite Review procedure. ,
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, ATTACllMI:NT I RI: PLY 10 NOllCE OF VIOLATION VIOLAllOfD50-156(457P07012-0M
- 3. 10 CFR 50, Appendix II, Criterion 111, " Design Control," requires that measures shall be established to assure that applicable regulatory requirements and the design basis, for those structures, systems, and components to which the appendix applies, are correctly translated into specifications, drawings, procedures, and instructions.
10 CFR 50, Appendix II, Criterion 111 also requires that design control measures shall provide for verifying or checking the adequacy of the design, such as by perfoimance of design reviews, by the use of alternate or simplified calculation methods, or by the performance of a suitable testing program.
Contrary to the above, on July 23,1997, design control measures established for checking the adequacy of calculation PSA-9513, Revision 0, "llyron/Braidwood Minimum Auxiliary Feedwater Flow for Feed Une lireak Analysis," failed to identify non-conservative and unrealistic flow resistance assumptions in the calculation.
IWASON FOR Tile VIOLATION in 1995, Nuclear Fuel Services (NFS) calculation P! A 119513 was performed. The individual who performed the calculation improperly used the esults from a Sargent and Lundy (S & L) calculation (MAD 90 0055) which was referenced. The S & L calculation determined the Auxiliary Feedwater (AFW) flow rates to each steam generator for various accident conditions.
This caleplation was based on theoretical ec:flicients of friction for the various system componerts (valves and piping). The individual who used the AFW flowrates detennined from the S & L calculation failed to review the actual calculation to detern:ine how the flowrates were calculated and also did not benchmark the results against actual AFW surveillance or test results.
Engineering procedure, NEP 12 03, " Nuclear Design information Transmittals (NDIT)," applies to the transmittal and review of the design documents between ditTerent departments or companies. The limitations and assumptions made with the S & L data in calculation PSA-Il 13 could have been identified with the proper use of this procedural guidance.
The NFS Engineer failed to follow the procedural guidance that had been established. This would have identified the limitations and assumptions made with the S & L data.
CORRECTIVE ACTIONS TAKEN AND RESULTS ACHIEVED A safety analysis comparison was performed that determined the non conservative error was fully bounded in the current design basis calculation used in the Feedwater Liae llreak (FLB) accident analysis, CN-TA 97-008, 9
A'ITACilMI:NT I R1 l'1310 NOTICII Of VIOLATlON y1QLATION f 50@f>f 4571'97012.(11)
Notes were inserted in calculations CN TA 97 008 and PSA II 9513 to direct the user that the flows specilled have been verilled to be conservative (in PSA ll 9710) and that the correct Al W flow should be used in any future calculations. Calculation PSA-II 9710 already ,
contained wording which infonned the user of this infbnnation.
ACTIONS TAKEN TO PitEVliNT llECUltillINCE A meeting was conducted with the Nuclear liuel Services Safety Analysis Staff to discuss the circumstances associated with the error that occurred in this situation to emphasize the irnportance of fbliowing the procedure requirements as well as supervisory oversight responsibilities. This meeting was led by the individual who had made the initial calculation error.
A review of a specific calculations perfonned by the Engineer who made the identified error was conducted by the Nuclear iluel Services Safety Analysis Supervisor. No additional errors were identified during this review. If problems had been noted, the scope of the review would have been expanded.
DATE WilEN FULL COMPLI ANCl! WAS ACillEVliD Full compliance was achieved when the calculation found to be in error was evaluated and results from this evaluation showed that the non-conservative error was fully bounded in the current design basis calculation, CN TA 97 008.
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l ATI ACllMI.NT I y-Id!'IN 10 NOllCI: Ol' VIOL.AllON
- Y10LA110N.(%4fy457r970lSO4) f
- 4. 10 CFR 50, Appendix 11 Criterion XVI, " Corrective Actions," requires, in part.
g that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and y(
equipment, and nonconformances are promptly id tified and corrected.
{s Contrary to the above, on July 22,1997, the team identified that since initial plam $
construction the Unit I refueling water storage tank (RWST) heater had not been mounted to the RWST tunnel floor in accordance with design drawings. As a result, the probability of a scismic event challenging the integrity of the Unit 1 RWST heating system and draining the RWST was increased.
ItEASON FOR Tlll! VIOLATION The concern associated with the Unit I refueling water storage tank (RWST) heater not being properly mounted was investigated and the cause could .mt be determined. A construction design change (ECN # 36722) had been issued, however it was not installed.
No modification was created to implement the construction design change following the initial fuel load.
CORRECTIVl! ACTIONS TAKEN AND Rl!SUlflS AClllEVED llwAp 33010. Attachment 11, was completed to evaluate the operability of the equipment and identify any compensatory actions.
The RWST heater was bolted to the floor as required. This was completed on September 19,1997.
ACTIONS TAKl!N TO PREVENT Rl!CURRIINCE Cunently, the design change process ensures that design changes are tracked from the time they are initiated through installation and close-out.
DATE WilEN FULI. COMPl.l ANCE WAS AClllEVED Full compliance was achieved when the RWST heater was bolted to the floor as originally designed, ti
[ A'TACllMI:NT I RI.I'IN 10 NOTICl;Of VIOLATION y10LAllON (50-456(457P9701S05) 5, 10 CFil 50 Appendix II, Criterion V, " Instructions, Procedures, and Drawings,"
- requires, b part, that activities affecting quality shall be prescribed by
, documented procedures of a type appropriate to the circumstances and shall be -
accomplished in accordance with those procedures.
> liraidwood Administrative Procedure (llwAP) 1250 2, " Problem identification and Investigation Procedure," llevision 5, dated July 2,1996, requires that a :
person identifying a problem that may affect the operability of plant systems or equipment shall immediately notify the shin engineer.
Contrary to the above, on February 21,1997, licensee personnel failed to infonn the shin engineer when a problem regarding 2A safety injection (SI) pump lube oil filter inlet pressure, which may have affected the operability of the 2A Si pump, was identified.
RIIASON FORIllii VIOLATloN On Friday, February 21,1997, an anomaly with the 2A safety injection (SI) pump oil filter inlet pressure relief valve was recognized by an lingineer during a review of ASMll surveillance results. Specifically, the suction and discharge pressures for the oil filter seemed higher than expected. The vendor r.anual was reviewed which made no reference to operation with oil pressures at higher than normal ranges. Attempts were made to contact the individual who had performed the surveillance three days earlier and the SI pump vendor to obtain more infonnation. Initial contact was made with both individuals on Monday, February 24, 1997. The individual who had perfonned the ,
surveillance indicated that no anomalies were identified during the actual surveillance.
The Si pump vendor did not have specific knowledge on the condition and needed additional time to investigate. The next day, the vendor indicated that the identified conditions were acceptable and the system could operate in with the higher oil pressures until the next outage when the proper repairs could be completed. It was concluded that an oil pressure control / relief valve was the cause of the higher oil pressure.
IlwAP 1250-2, " Problem identification and Investigation Procedure" (itevision 5) includes a note which states the following:"Immediately notify the Shin Fngincer if the event or condition may be reportable to a llegulatory Agency, may affect the operability of plant systems or equipment, or presents a danger to personnel safety." At the time the high eil pressure was recognized, the individual recognized that the noted oil pressure difTered from the range specified in the vendor manual and investigated the condition before documenting the issue on a Problem !dentification Fonn (PIF). The individual should have communicated the identified concern to the Shin in a more timely manner, 12
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, A1TACllMl:NT 1 f(1:Pl,Y 10 NOTICli Of V101.ATlON ylOLA110N (50 456(437l'97012-01)
CORRECTIVl! ACTIONS TAKEN AND Rl!SULTS ACillEVED
'lhe individual who failed to notify the Shift of his equipment concern was appropriately counseled.
Following the initial investigation of the higher than normal oil pressures, it was concluded that the condition was in fact not an operability concern.
In July of 1997 Operability Assessment # 97-096 was perfonned which confinned that the 2A SI pump was operable with the higher than nonnal oil pressure condition and the pump was capable of perfbrming its design function during a design basis accident in that condition.
The malfunctioning relief valve which caused the higher oil pressure condition was repaired during refueling outage A2R06.
ACTIONS TAKEN TO PREVENT RECURRENCE The need for prompt communications with Operations regarding abnormal equipment perfbnnance was emphasized with Engineering personnel. The requirements stated in the station's current PlF Procedure, NSWP A IS, " Comed Nuclear Division Integrated Reporting Program," were discussed during an Engineering Department all hands meeting.
DATE WilEN FULI COMPl LANCE WAS ACillEVED Full compliance was achieved when the Shift manager was notified of the equipment concern and the individual who failed to notify the Shift of his concern in a timely manner was appropriately counseled.
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[ ~* NITACllMI:NT l Iti:I'IN TO NOTICI: Ol' VIDIAllON Y101ADON f 50-45.6(45719701504)
- 6. 10 CFit 50, Appendix H, Criterion XV, " Nonconforming hiaterials, Pans or Components," requires, in part, that measures shall be established to control material, parts, or components which do not confbrm to requirements in order to prevent their inadvertent use or installation. These measures shall include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification so afTected organizations.
Contrary to the above, on February 18, 1997, licensee measures failed to effectively control and prevent the installation of a non-environmentally qualified breaker in a harsh environment. Specifically, the licensee installed a non-environmentally qualified circuit breaker in the 1 A residual heat removal pump minimum flow valve circuit breaker cubicle, which was considered to be a harsh environment.
JtIIASON 1 OR Tilli VIOLATION As a result of a unintentional trip of a motor operated valve (hiOV)in 1995 at Dresden, liraldwood performed a review of the trip settings of all safety related h10V circuit breakers. During this review,28 hiOVs were identified as requiring an increase in their breaker trip seipoints to prevent the nuisance trips in the unlikely event of a motor reversal. For the most part, these changes were accomplished by adjusting the instantaneous setting on the molded case circuit breaker. In case of 2 h10Vs, (11111610 and 1111161!), it was necessary to change out the breakers since the installed models did not have adjustable instantaneous trips.
lireaker 11(11610 is located in hiotor Control Center (h1CC) l AP211? which is in a harsh I?Q cnvironment. This breaker was changed out under Setpoint/ Scaling Change itequest (SScit)95-049. The SScit specifically stated that Stores item Number, (SIN),767C26 should be used as the replacement breaker. This SIN was for a non liq qualified breaker and should not have been used Ihr this application. The individual who specified this Stores item Number overlooked the fact that the breaker was to be located in a harsh !!Q environment. This error was also overlooked by the work analyst who failed to recognize the problem during an independent review and the Quality Control (QC) inspector who failed to recognize the problem during a review of the work package. The probable cause of these missed opportunities is due to the fact that engineering specifically identified the SIN for the replacement breaker.
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A1TACitMENT I REPIN 10 NO11CE Ol' VIOLATION y10LA110N (50 43.g4571'97012 06)
CORRECTIVE ACTIONS TAKEN AND RESUI TS ACllll?VED An operability determination (97 098) was conducted to evaluate the breaker. It was concluded that the installed breaker would be able to perform its safety function.
The non EQ qualified breaker for 1Ril610 was replaced with an appropriately qualified
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breaker on August 26,1997.
ACTIONS TAKEN TO PREVENT RECURRENCE Training was held with Design Engineering personnel to discuss the details of this event, as well as the need to have the hiaterials hianagement Group perform a parts evaluation prior to specifying material by Stores item Number.
The event details were also communicated to work analysts. In addition, the need to independently determine the proper Stores item Number Ibr parts was communicated to these individuals.
Quality Control Inspectors also discussed the event as well as the need to question mismatches between the EQ classification of Nuclear Work Requests (NWRs) and material red tags.
Appropriate maintenance personnel were tailgated on the importance of questioning the installation of non EQ equipment under EQ classified Nuclear Work Requests.
DATE WilEN Full COMPl 1ANCE WAS AClllEVED Full compliance was achieved when the non EQ qualified breaker was replaced with an appropriately qualified breaker.
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