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| {{Adams | | {{Adams |
| | number = ML20154G864 | | | number = ML20205C233 |
| | issue date = 09/16/1988 | | | issue date = 10/19/1988 |
| | title = Insp Rept 50-482/88-200 on 880606-17.Potential Enforcement Findings & Observations Noted.Major Areas Inspected:Line Organization Support & Contribution to Plant Quality & Quality Verification Organization Abilities Re Deficiencies | | | title = Discusses Quality Verification Function Insp Rept 50-482/88-200 on 880916 & Forwards Notice of Violation. Preparation to Discuss Potential Enforcement Finding Noted in Rept & Item from Earlier Insp Rept at Meeting Requested |
| | author name = Correia R, Finkel A, Hawkins F, Hopkins P, Hunter D, Moore R, Prescott P, Sparks S, Weiss S | | | author name = Callan L |
| | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV), NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR) | | | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| | addressee name = | | | addressee name = Withers B |
| | addressee affiliation = | | | addressee affiliation = WOLF CREEK NUCLEAR OPERATING CORP. |
| | docket = 05000482 | | | docket = 05000482 |
| | license number = | | | license number = |
| | contact person = | | | contact person = |
| | document report number = 50-482-88-200, NUDOCS 8809210095 | | | document report number = NUDOCS 8810260427 |
| | package number = ML20154G849 | | | package number = ML20205C239 |
| | document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | | | document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE |
| | page count = 26 | | | page count = 2 |
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| Enclosure 3
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| U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION Report No. 50-482/88-200 Docket No. 50-482 License No. NPF-42 Licensee: Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, Kansas 66839 Facilit Wolf Creek Generating Station .
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| Inspection At: Wolf Creek, Burlington, Xansas, June 6-17, 1988
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| b Richard P. Correia, Senior Operations Engineer T!h!fd (Date) i NRR(TeamLeader)
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| tedh Perry C. Hopkins, Res'ident Inspector 9b6k]
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| / (Dite)
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| Region !!
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| S(RMik Randolph L. Moore, Reactor Inspector vHn
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| Region II
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| b, 9 88 Scott E. Sparks, Reactor Inspector / (Date) c Region II SYUN k Dorwin R. Hunter, Senior Reactor Inspector 4/n
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| Region IV d 9; CfS$
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| Allen E'. Finkel, Reactdr Inspector / (Date7
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| Peter J. Prescott, Qudlity Operations ' Engineer / (Date)
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| NRR Reviewed by: I (
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| CDTikins, Chief, Qu111ty Operations Section /(Dite)
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| i NRR ,
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| Approved by: . 7/(!80 5. H. Weiss, Chter, Quality Assurance Branch /(Da%J NRR ggj92h$ch b .$
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| SUMMARY d
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| Areas Inspected This special, announced Nuclear Regulatory Comission (NRC) team inspection was the seventh in a series of NRC Headqttarters-directed Quality Yerification Function Inspections (QVFIs). The inspection was performed to assess the line organization's support and contribution to plent quality and the quality verification organization's ability to identify, solve, and prevent the occur-renct 4 safety-significant deficiencies in the functional areas of plant
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| a, is and maintenance. Another area that was evaluated during the QVFI he effectiveness of management in ensuring that identified quality j- t .. fencies were responded to promptly and completely.
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| l Results Within the functional areas of operations and maintenance, six potential enforcement findings (pEFs) were identified: (1) six examnles of not taking appropriate corrective actions to prevent recurrence of plant system and l component deficiencies, (2) not having procedures and instructions appropriate
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| for the bearing removal activities on a component cooling water pump, (3) not obtaining and performing evaluations of applicable service information letters from the emergancy diesel generator vendor, (4) not verifying that four seismic and vibration control supports were installed on the emergency diesel generator turbocharger cooling water piping as specified by the vendor's design drawing, (5) not posting a fire watch after a fire barrier seal.in a penetration was determined to be unqualifier', and (6) not declaring a loop of the Essential Service Water System inoperable when it '.4s determined it did not meet its :
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| specified design requirement In addition, two observations were identified: (1) a lack of a feedback mechanism for maintenance personnel to report problems and recomendations to ,
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| procedure write n , and (2) a lack of an adequate methodology to calibrate the resistance temperature detectors for the reactor coolant syste .
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| e 1 INTRODUCTION This special, announced NRC team inspection at Wolf Creek Generating Station (WCGS) sas performed to evaluate the acceptability of the line and quality verification organizations' activities and management's support of these activities. The inspection was the seventh in a series of NRC headquarters-directed inspections performed.under the guidance of NRC Inspection Manual Temporary Instruction 2515/78, "Inspection of Quality Verification Functions."
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| The inspection consists of personnel interviews, direct observation of in-process activities, and review of work document Quality Verification Function Inspections (QVFIs) are not intended to verify licensee compliance to administrative controls; they are intended to verify the technical adequacy of safety-related activities. However, if deficiencies are tvund in these activities, the underlying procedures and adminictrati';e con-trols are reviewed. The intent of these inspections is to improve plant opera-tional safety through inspection processes that are focused on activities that affect plant safety and reliabilit The QVFI at Wolf Creek focused on plant operations and maintenance of plant systems ud components. The inspectors reviewed selective samples in these and closely associated areas to identify safety-significant problems to be used as the vehicles for evaluating the effectiveness of quality achievement and verification. The results of this review are discussed below and the inspectors' more significant findings are categorized as potential enforcement tind!ngs and observation Potential enforcement findings are apparent violations of regulatory require-ments that will be further evaluated by NRC Region IV management for possible enforcement action. Observations are items that may not violate any regulatory requirements and may not violate plant procedures, but that appear to be less than optimum. Observations are being referred to NRC Region IV and NRC Headquarters Staff and may require inspectors to perform followup reviews during subsequent inspection PLANT OPERATIONS 2.1 Control Roon and Operations Activities 2. Inspection Results The NRC inspectors observed control room and other eractions activities, interviewed control room personnel, and reviewed rertinent documents related to operations activities. The inspectors observed control room decorum, control rocm shif t turnover during dayshift and backshift, main turbine valve cycling, mainte,ance and testing of the reactor trip breakers, a valkdown of the auxili-ary feedwster system, and a transient involving a loss of automatic feedwater control and the subsequent recovery in the control room. The team inspected other plant areas to verify operability of equipment, control of ignit sources and combustible materials, proper condition of fire detection extinguishing equipment, adequacy of maintenance activities, and adequa of selected surveillance '
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| .. | | In Reply Refer To: OCT I 91988 Docket: STN 50-482/88-200 Wolf Creek Nuc7 ear Operating Corporation ATTN: Bart D. Withers President and Chief Executive Officer P.O. Box 411 Burlington, Xansas 66839 Gentlemen: |
| Control room shift turnovers were orderly and briefings of individual operators were adequate. The NRC inspectors observed that the oncoming shift conducted another briefing for all operators after the off-going shift had left the -
| | This refers to the Quality Verification Function Inspection (QVFI) team inspection conducted during the period June 6-17, 1988. The inspection team findings were documented in NRC Inspection Report 50-482/88-200, dated September 16, 198 Following further evaluation of the potential enforcement findings (PEFs) |
| control room. During these briefings (generally less than 5 minutes in dura-tion), the operators discussed scheduled surveillance testing and general plant status information. During the QVFI, the plant experienced a loss of automatic feedwater control that led to a system transiant. The operators quickly assessed the plant condition and responded to avoid a reactor trip on a low steam generator water level. The NRC inspectors observed that during the shift turnover briefing attentive after the to the briefing transient, incoming informatio shift operators The licensee's were very(QA)
| | ide.itified in the NRC inspection report, it was determined that certain of your - |
| Quality Assurance organization has audited this area several times and has not identified any problems with the adequacy and effectiveness of shift turnover The NRC inspectors noted on several occasions that the operations manager and plant manager were in the control room observing shift turnover activities, other plant evolutions, and the shift supervisor's activities durinq differ-ent evolutions. At most times, an additional senior reactor operator (SRO) was available during the day chift. There also was good administrative-clerical support for the supervisors and operating staff. These support personnel appeared to remove some of the administrative burdens from the control room staf Management appeared to support quality operations and responded well to operators' recommendations concerning the use of operator aids in the control room. For example, a suggested operator aid, which consisted of a magnetized plastic card inscribed with the technical specification requirement and limit-ing condition of operation (LCO), was used on the engineered safety features actuation system bypass panel. This magnetized cara covered the bypass key lock. When a system channel had to be bypassed, the operator aid had to be removed before inserting the key. The card then was placed in front of the control room operator to serve as a constant reminder of the condition that had to be monitore Managecent also has provided opportunity for operators to part'cipate in a
| | activities were in violation of NRC requirements. Consequently, you are required to respond to these violations, in writing, in accordance with the provisions of Section 2.201 of the NRC's "Rules of Practice," Part 2 Title 10 Code of Federal Regulations. Your response should be based on the specifics contained in the Notice of Violation enclosed with this lette One of the potential enforcement findings of the QVFI report (Item No. 88-200-8) is not included in the appended Notice of Violatio . |
| .ollege training program. These personnel are . ant to a local university to gain college credit towards meeting qualification requirements for a shift .
| | Additionally, with respect to potential enforcement finding 88-200-1a, it is stated in the QVFI report that the issue is also addressed in NRC Region IV Inspection Report 50-482/88-19 and will be followed up by Region IV. You are requested to be prepared to discuss all aspects of these two items with Region IV staff at a fu'ure meeting. You will be contacted shortly to arrange a mutually acceptable tirre for the meetin The response directed by this letter and the accompanying Notice is not subject ' |
| ' technical advisor (STA) position. There were times when several SR0s on the l same shift had the qualifications of an STA, which provided extra crs of i technical expertise to evaluate specific plant problems. The NRC .mpectors l observed that this program appeared to create higher morale and lower person-nel turnove lne NRC inspectors observed plant operator surveillance activities of technical l specification requirements for safety-related systems and components. The inspectors also observed the licensee's GA overview of these operator surveil-lance activities. The QA personnel who were observed provided effective identi-fication of problem areas during their overview. The NRC inspectors observed portions of 22 selected surveillance tests aru all aspects of several other tests. Qualified personnel performed the tests and properly calibrated required test instrumentation, and the resulting data met the requirements of the Tech-nical Specifications. When discrepancies were identified, they were rectified and the systems were properly returned to service. QA personnel were present while NRC inspectors observed surveillance test . _ _ _ _
| | to the Paperwork Reduction Act of 1380, PL 96-51 Should you have any questions concerning this inspection, we will be pleased to discuss them with yo ' |
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| The NRC inspectors watched tag out and equipment restoration on several occasions.. The tag out and restoration processes, including briefings, were well understood bv all operators who were involved. The NRC inspectors .
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| observed that QA personnel regularly reported and followed up on fintiings in these area The NRC inspectors also watched operators pwform a cycling test of the main turbine valves in accordance with Procedure STS AC-001, Revision 5. This surveillance-test demonstrates the operability of the turbine overspeed pro-tection system as required by Technical Specifications. The operators compe-tently perforaed the test, and they adhered closely to the procedure. The test was satisfactorily completed without any irregularities or component mal-function The NRC inspectors performed an auxiliary feedwater (AFW) system walkdown with a reactor operator using Procedure CKL AL-120, Revision 10, "Auxiliary Feedwater Normal Lineup," and piping and instrumentation diagram (P&ID) drawir.g M12AL01(Q),
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| Revision 0. The NRC inspectors determined that the actual system configuration agreed with the P&ID drawing and that the operator appeared knnwledgeable of valve locations and proper valve positions. The valves were found to be free of corrosion, locked if required, and positioned in agreement with the P&ID and the procedure. During the walkdown, the NRC 1:. pectors identified fotrvalves(EF-V07/,FC-V115,AB-V085,GF-V009)withnolabelsandonevalve (AL-V035) that contained a small packing leak. The reactor operator noted all deficiencies and they were corrected after being discussed with plant managemen During the AFW turbine-driven AFWwalkdown,(the pump TDAFWP) NRC speedinspectors set point on noted that theshutdown the auxiliary position of the panel did not agree with required TDAFWP set point noted in Procedure CXL AL-12 The required speed set point was 3850 rpm, while the actual control set point was 5750 rpm. The inspectors discussed this discrepancy with cogni-zant instrumentation and control (I&C) personnel, who explained that the TDAFWP shutduwn parel controller output signal to the pump is 3850 rpm, regardless of the higher set point. 1&C personnel also provided, as a verification of this condition, the results of the testing of the TDAFWP controller conducted on j October 23, 1987, in accordarce with Prucedure INC L-1000, Revision 2. The NRC inspectors discussed the TDAFWP speed set point with several operations personnel and determined that operator knowledge of equipment operation was i Meeptabl .1.2 Results Sumary The NkC inspectors observed during multiple dayshifts and backshifts that
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| control room operators conducted themselves in a professional manner.
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| ; Operators appeared to be attentive, were knowledgeable of plant status, and performed testing correctly with close adherence to procedures. The NRC l
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| inspectors verified that QA personnel did observe performance of severel i operational and maintenance work activities. The NRC inspectors' observations I relating to the shift turnover briefings emphasize the need for operator atten-l tiveness at all time . - - __ _ . _
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| 2.2 Quality Assurance and Control Activities 2. Inspection Results
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| The NRC inspectors observed 0A activities, interviewed QA personnel, and l reviewed applicable QA aud) ind surveillance reports in the operations are Specifically, the inspectors abserved quality control (QC) (a part of Quality Department at Wolf Creek) involvement in maintenance and testing of reacto" trip breakers and QA personnel performing a followup surveillance. The NRC inspectors also reviewed audits and surveillance reports that involved opera- l tions activities, as well as those covering general work contro QA personnel audited noimal, backshift, and weekend activities and surveyed operations activitie The NRC in @cctors observed that QA personnel were knowledgeable and competent in the V audits and curveillances and maintained an adequate mixture of direct QA uservation of operational activities and review of documentatio The NRC inspectors reviewed audit reports that spanned approximately 2 year The quality of the reports and types of observations had recently improved, covering more of the actual performance of the activity rather than verifying strict compliance to procedures. An essential elements book was written and implemented by the QA department to ensure tnat the essential elements of test procedures were critically analyzed by a QA auditor during his or her verifica-tion activitie The NRC inspectors observed that during the performance of maintenance on the reactor trip breakers (Work Request 50762-88), QC personnel were present and verified the completion of several in-process inspection hold point In addi- '
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| tion, the NRC inspectors acccmaanied a QA inspector during a followup sur-veillance of plant equipment, )oth safety related and nonsafety related, and of general plant conditions. This surveillance was performed to verify that corrective actions for deficiencies identified in QA audit TE53359 S-1627,
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| "Control of Plant Equipment," had been implemente During the followup surveillance, the licensee's QA inspector identified several unacceptable conditions, including one that involved the storage of safety-related snubbers in the auxiliary building. More specifically, approxi-mately nine mechanical snubbers had been functionally tested in early May 1988 and three had failed. Although all nine of the snubbers were appropriately tagged, the licensee did not segregate the failed snubbers from the snubbers that passed testing. In addition, all nine snubbers were stacked together 19 an area not designated for storage. When the NRC inspector questioned the acceptability and adequacy of this condition, licensee management had the snubbers moved to a proper storage area used for safety-related equipment. The licensee's QA inspector also identified a leaking valve on the second stage feedwater reheater drain tank. This valve, AFV 944, is a level switch isola-tion valve, and it contained a body-to-bonnet steam leak. The QA inspector reported this condition and subsequently the valve was repaired with furmanite to stop the leak. It was apparent to the NRC inspectors that the QA inspector was knowledgeable of proper plant conditions and of the need to promptly report result, to managemen The NRC inspectors also reviewed approximately 10 recent QA audit and sur-veillance reports. One of these reports, QA Audit TE50140 K-192, "Corrective
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| Actions," identified 0-rings in the solenoid oserators of post-accident sampling system containment isolation valves tut were not environmentally qualified (EQ). Licensee personnel discovered the 0-rings that were not EQ in November 1987 during the implementation of a plant modification request (PMR1844). This moJification involved changing valve solenoid springs in several valves. During the implementation of PMR 1844, a maintenance crew mistakenly disassembled the solenoid operator of a valve (GS-HV-013) not requiring modification. The crew realized t!.eir error, and they also identi-fied that the solenoid operator contained EPR-type 0-rings that were not EQ for that specific application. Corrective work requests were written to inspect the solenoid operators and to replace the EPR 0-rings that were not EQ with EQ grafoil 0-rings, as necessary. The following valves were inspected: con-tainment hydrogen control valves GS-HV-4, GS-HV-C GS-HV-9, SS-HV-13, GS-HV-14, and GS-HV-10; nuclear sampling valves SJ-HV 3, SJ-HV-4, SJ-HV-5, and SJ-HV-128; and steam generator blowdown valves BM-HV-35, BM-HV-06. BM-HV-37, and BM-HV-?a, The QA organization issued a defect / deficiency report (D/DR 87-132) after discovering that the 0-rings in the valves were not EQ. The QA organization also issued a quality plant deviation (QPD) and a programmatic deficiency report (PDR OP87-111) to address the disassembly of the wrong solenoid operator during implementation of PMR 1844. An engineering evaluation (87-SJ-10) was performed to determine the effect of having the 0-rings that were not EQ in the valves. The results of the engineering evaluation showed that moisture or water that might intrude into the solenoid operator if an 0-ring that was not EQ failed would not affect the valve's pressure retaining function; however, moisture could cause the valve to remain in the failed-closed position upon receipt of a containment isolation signal and not allow the valve to reopen to operate the post-accident sampling system. All of the work requests for the affected valves had been completed at the time of the engineering evaluatio It could not be determined how many of the 14 valves in question had contained 0-rings that were not EQ because the 0-rings in all valves were changed and the licensee did not docu'nent which of the valves had the 0-rings that were not E The NRC inspectors determined that the valves were originally delivered with EPR 0-rings that were not EQ, but were subsequentl Package (OCP) CS-90-W, Field Change Work Request (y modified by Design Change work permits CWP BM-212-E, CWP-GS-651 and work request WR698-85. At the time of the QVFI, it was unclear whether the 0-rings had actually been replaced during implementation of the work permits and request or whether additional work on solenoid operator 0-rings had been performed on the valves after the original issuance of the CWPs and WR. However, it is apparent that during the time when FJ 603A-02 was issued and CWP BM-212-E, CWP-GS-651 and WR 698-85 were all com-pleted by January 21, 1985, the maintenance organi::ation did not adequately accomplish the specified activities and the QC organization failed to verify, during their reviews and inspections, that the proper EQ 0-rings had been installed. The licensee's actions to ensure that the deficient valve operators had EQ 0-rings installed corrected the immediate problem. However, the NRC inspectors detemined that the licensee had not investigated the underlying cause which permitted installation of 0-rings that were not EQ to remain installed in the solenoid operators. This failure to determine the underlying cause of the conditiN is considered a potential enforcement finding (Item No. 88-200-1a). This issue is also addressed in NRC RIV Inspection Report 50-482/88-19 and will be followed up by Region I \ - ___ , - - _ _ .
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| I 2.2.2 Results Sumary l The NRC inspectors detennined that QA activities generally were conducted in a performance-oriented manner by qualified individual .3 Operations Training 2. Inspaction Results The NRC inspectors reviewed licensed, non-licensed, and craft training practices. The NRC inspectors' interviews with instructors indicated that the instructors were competent and professionally trained. Instructor performance is evaluated by the manager of training as well as by seer, self, technical peer, and supervisory personne Each instructor had )een appropriately certified for the activities he or she was performing. There currently are four positions for licensed instructors, two were filled by qualified contract personnel, and two were vacant. The training staff and the instructional staff appeared to be dedicated, professionally competent, and responsive to student concerns and need .3.2 Results Sumury The NRC inspectors were concerned that two vacancies in positions for licensed instructors exists in the licensee's training department. This issue was discussed with licensee management to emphasize the importance of training and the need for a fully staffed training departmen PLANT MAINTENANCE 3.1 Maintenance Activities 3. Inspection Results '
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| The NRC inspectors observed maintenance activities on a pressurizer code safety valve and a component cooling pump and evaluated the engineering support activities for maintenance on a pressurizer spray valve. The inspectors reviewed the following attributes of each maintenance activity: quality of instructions and worker training, familiarity of worker with the task and with ~
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| tools and equipment, listing of task precautions, adherence to procedures, and QC involvement in the activit .1. Pressurizer Spray Valve The NRC inspectors reviewed engineering calculations generated by Nuclear Plant Engineering personnel in support of the encapsulation of a pressurizer spray valve packing box. The encapsulation was necessary to control a reactor cool-ant leak from the packing box assembly. The NRC inspectors determined that the calculations were detailed and accurate. Engineering personnel performed a thorough analysis that demonstrated good support of this maintenance activit .1.1.2 Pressurizer Code Safety Valve The NRC inspectors watched the maintenance technician set up and clean the components on a pressurizer code safety valve in preparation for disassembly and rework. The work was well organized and managed. The NRC inspectors also
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| reviewed the applicable maintenance procedure to be used for this activity and determined that the detail, references, precautions, tool requiremer. s, and other important data were adequat .1.1.3 Component Cooling Water Pump Disassembly The disassembly of the component cooling water (CCW) pump was a relatively com-plex task that relied hea/ily on skill-of-the-craf The work instruction con-sisted of six general steps on the work request form and a reference to an attached photocopy of a section of the pump vendor's manual. The procedure used for disassembly was also photocopied from the pump vendor's manua During the work to remove the pump's bearing, the NRC inspectors observed that the maintenance technician w6s using a bearing puller on the bearing while heating the bearing housing with a gas flame torch. The technician involved was knowledgeable of the process, but not of potential effects that heating might have on the material characteristics of the bearing and pump shaft. No method was .specified, nor was a contact thormometer on hand to determine the temperature of the heated parts. The NRC inspectors noted that the instruction to remove the bearing simply stated "remove the bearing." The work instruction did not include a caution statement addressing the potentir' damage to the pump shaft or bearing, heating instructions, expected temperature for bearing release, or maximum temperature recomendation QC inspectors were not present during this activity because it was not con-sidered a detailed step requiring a QC hold point. Apparently, the Quality organization responsible for procedural reviews determined that this bearing removal did not require additional details and that skill-of-the-craft was adequat The NRC inspectors discussed the lack of temperature limits and a heating process description and control with the procedure writing group in Maintenance Engineering. In response, the engineering supervisor stopped maintenance activities until the pump vendor could be consulted. Following consultation with the vendor, a bearing surface temperature limit of 750'F was specified, an
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| expected bearing release temperature of 300' to 500'F was established, and heating process instructions were provided in a revision to the work reques The revision also indicated methods for monitoring bearing and shaft temperature This CCW pump bearing removal activity indicated a weakness in the work process with regard to the appropriatness and adequacy of crocedures and work instructions and is considered a potential enforcement finding '(Item No. 88-200-2).
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| Additionally, the assumptiun that skill-of-the-craft was sufficient for this i activity was not prudent and demonstrated poor comunication between procedure writers and task performers (Observation Item No. 88-200-3).
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| 3.1.2 Results Su mary The NRC inspectors concluded that general maintenance technician performance was good, QC presence during performance was adequate, mairtenance craft knowledge and experience levels were adequate, but work instructions, especially with regard to limitations and precautions, were weak.
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| The NRC inspectors determined that the probable causes of the v:ork instruction weaknesses were the informality of work inst uctions, unfamiliarity of pro-cedure writers with the task to be performed, inadequate attention to detail, .
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| and a lack of feedback from maintenance personnel to cognizant enginesrs on problems they encounter and recomendations to improve the instruction .2 Control Building Heating, Ventilating, and Air Conditioning (HVAC) System 3. Inspection Results From 1985 until now, the control room ventilation isolation signal (CRVIS)
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| system has been activated 72 times as a result of spurious signals from the chlorine monitor system, and radiation detectors and other components in the HVAC system. More specifically, 28 of the CRVIS actuations have been attributed to malfunctions of the chlorine; monitor system and the remaining 44 to problems with the radiation detection system, electrical circuit breakers and dampers within the HVAC system. The following sections detail the inspector's review of the three opparent contributors to the CRVIS actuation .2.1.1 HVAC Breakers The NRC inspectors reviewed records pertaining to problems with the control building HVAC circuit breakers. In early 1985, the licensee's Maintenance Engineering organization identified nuisance tripping of the HU-B100-0501 ITE b/eakers at their respective motor control centers (MCC). An engineering evaluation request (EER 85-GK-08) was prepared by Maintenance engineering on July 2ti, 1985. The resulting engineering evaluation, complated on November 27, 1985, stated that new breakers would be ordered with a specified instantaneous trip settin The NRC inspectors determined that the licensee had received the breakers ordered by engineering, but had never installed them in the designa+ed syste Since the maintenance organization was not notified that the brea; vs had been received, their work request records indicated that this item was open because the parts were not available. The NRC inspectors determined that of the two breakers ordered for this system, one was in the warehouse and the other had been used in another system and not installed into the appropriate MCC as specified in engineering disposition REDA 0-E-1324-GK. It appeared that no one was tracking this item to ensure that the replacement breakers were installed as directed by engineering. This failure to take the specified corrective actions regarding the HVAC electrical system breakers malfunctions is considered a potential enforcement finding (Item No. 88-200-lb). In response to this issue, the licensee has committed to evaluate the existing engineering evaluation request tracking syste .2.1.2 HVAC Dampers The inspectors reviewed records pertaining to problems with the control build-ing HVAC dampers. The records indk:ated that during routine work, maintenance engineering personnel found that the HVAC dampers were not aligned as required by the design drawing. Although Maintenance Engineering determined that the observed misalignment was the cause of the damper failures, it was not evident whether Maintenance Engineering considered the cause of the misalignment during the investigation of the damper problem. Additionally, the investigation into the cause of the failures did not consider whether the multiple CRVIS actuations
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| also were contributing to the damper problem. These failures to fully investigate the underlying causes of the multiple HVAC damper ftilures is considered a potential enforcement finding (Item No. 88-200-1c). .
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| 3.2.1.3 Control Room Habitability System Chlorine Monitors !
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| The NRC inspectors reviewed the specification for the replacement chlorine detector monitors that are part of the control room habitability systein and verified that the site-suecific technical requirements for the monitors were defined within the specification criteria. The NRC inspectors also reviewed the engineering design calculations to ensure that the technical specification requirements were considered when evaluating the new design criteri The chlorine monitors are essential elements of the control room habitability systems. These habitability systems permit access to and occupancy of the control room during normal plant operations as well as during and following emergency conditions. They also are designed to enable the plant operators to achieve and maintain the plant in a safe shutdown condition following a design-basisaccident(DBA)
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| As discussed previously, the chlorine monitors have caused 28 actuations of the control room ventilation isolation signal system (CRVIS) since 1985. Eighteen of these actuations were due to paper tape problems, seven were due to signal spikes from the chlorine monitors, and three were attributed to causes such as manual actuation and personnel errors. Operations personnel currently are required to survey the chlorine monitors twice per shift to look for indica-tions of a possible malfunctio The NRC inspectors reviewed a recent engineering study that had been conducted to provide solutions to prevent further malfunction of the control room chlor-ine monitoring system. This study indicated that the major problems with the chlorine monitors were tape failures, electrical failuras, spurious spikes with tape failures, and failures of lamps. The licensee recently issued a work order to remove a WISA puma from its present location in the chlorine monitor unit to a remote location )ecause the licensee believed that WISA pump vibra-tions may have been contributing to the problems. The results of this modifi-cation will not be known until sufficient operating time has elapsed.
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| l The licensee also plans to replace the 7040 MDA model monitor with a commercial l grade Delta chlorine detector system during the next outage scheduled for the
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| : lest quarter of 1988. The Delta system is to be dedicated and qualified during l the third quarter of 1988. In addition, the licensee has ordered a Sensidyne chlorine detector system to back up the Delta chlorine detector system. The i
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| NRC inspectors determined that the licensee's activities to replace the present MDA chlorine monitor system with Delta and Sensidyne systems were positive actions to resolve the problem.
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| l The liceqsee has experienced a large number of CRVIS actuations resulting from
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| the cMorine monitoring system malfunctions without aggressively pursuing resolution of the problem until recently. Because of the large number of CRVIS actuations attributed to chlorine monitor malfunctions since 1985 and the
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| , apparent slowness with which the licensee has taken action to correct the l problem, this matter is considered a potential enforcement finding (item No.
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| I 88-200-1d).
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| 3.2.2 Results Summary On the basis of the above, the licensee's program for determining the under- -
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| lying causes of plant system and component failures and malfunctions needs strengthening. The fragmentation of responsibility for implementing the WCGS corrective action program appears to be contributing to the program's weaknes With the exception of maintenance technicians, no single organization has been given the responsibility to technically analyze failures and malfunctions to determine their underlying causes. Additionally, the licensee's investiga-tions of component failures and malfunctions do not always consider their effect on interrelated system Without adequate cause evaluation information, thorough analysis of failures and malfunctions cannot be made and the trending programs become merely failure frequency indicators. Trending information should be used to increase the reliability of plant systems through early detection of repetitive component failure .3 Maintenance Measuring and Test Equipment 3. Inspection Results The NRC inspectors reviewed the QA activities associated with measuring and testequipment(M&TE). The inspectors selected a review sample of M&TE used on various maintenance activities to determine the adequacy of the out-of-tolerance evaluations, of the historical documentation of M&TE use (use-history), and of the QA corrective action process. Additionally, the inspectors reviewed 10 randomly-selected out-of-tolerance evaluations to verify timeliness and tech-nical adequac The NRC inspectors determined that, with one exception, out-of-tolerance evaluations were performed in a timely manner and were technically adequat That exception, an evaluation for micrometer No. WC-6710, indicated that past usage of the lost instrument was acceptable because the previous two annual calibrations were within acceptable tolerances. In this case, better assurance of the microreter's accuracy during previous use would have been provided by a remeasurement of affected activities to verify if the previously taken meaure-ments were within expected ranges. The inspectors considered the micrometer example to be isolate . Results Summary The NRC inspectors determined that the measuring and test equipment program adequately supports ongoing maintenance activitie .4 Fire Protection System 3. Inspection Results The licensee has experienced a high instance of alarms activating as a result of the malfunction of a specific type of microswitch u:;ed in the fire protec-
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| tion system. These microswitches are installed on various outdoor valves that are located above and below ground. The Maintenance Engineering organization issued EER 87-FR-06 on May 8, 1987, which stated that the present micro-switches, Type PV IS-B, are routinely found corroded and are being used in
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| ' applications for which they were not designed. The NRC inspectors reviewed 16 recent work requests associated with microswitch failures and found that the microswitches continue to be misapplied. At the time of this inspection, the-licensee had not taken action to stop using the microswitches in applications for which they were not designed. This failure to take actions to resolve the apparent misap ment finding (plicaton of the microswitches is considered a potential enforce-ItemNo.88-200-le).
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| 3.4.2 Results Summary The NRC inspectors determined that the fire protection system was adequat However, the inspectors were concerned that the control room alarms that resulted from the malfunctioning microswitches may desensitize the operatnrs to an actual fire protection system actuatio .5 Emergency Diesel Generator Vendor Service Information Letters 3. Inspection Results The NRC inspectors reviewed several service information letters (SIls) to determine whether proper evaluation and implementation of any necessary component inspections and modifications had been performed by the license These SIls were issued by the emergency diesel generator (EDG) vendor, Colt Industries, to convey vital service information to its customer During an interview, licensee personnel told the NRC inspectors that Colt SIls are considered vendor technical information, which is to be reviewed and evalu-ated under Wolf Creek's Industry Technical Information Program (ITIP). The ITIP was established in response to NRC Generic Letter 83-28, Section 2. However, when the NRC inspectors asked to review the evaluation of Colt SILs conducted under the ITIP, licensee personnel gave the NRC inspectors an inter-office memorandum (No. AD 87-0373) dated November 9, 1987, which stated that no Colt SILs had been transmitted to ITIP personnel for their review and evalua-tion because of miscommunications between the vendor (Colt), the plant's architect-engineer (Bechtel), and licensee personnel. The memorandum also requested that Colt be contacted to determine which SILs were applicable to Wolf Creek and to send them for immediate review by ITIP personne Colt determined that there were five SILs that pertained to the EDGs supplied to the licensee. Licensee personnel stated that they had received the SIls from .
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| Colt in January 1988. However, at the time of this inspection, the NRC inspec-tors found no formal review or evaluation of the five SILs had been performed by the licensee. Further, the inspectors determined that the licensee had not received three other SILs that pertained to the Wolf Creek EDGs. This failure to obtain all relevant Colt Sils, review them to determine their applicability to WCGS, and evaluate their relevance to the Wolf Creek EDGs is considered to ,
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| be a potential enforcement finding (Item No. 88-200-4). Subsequent to this finding, the licensee issued Programmatic Deficiency Report OP-88-12a and issued an engineering evaluaticn request to determine if additional Colt SIls, which were applicable to Wolf Creek, existed and had not been received by ITIP personne _ _ _ . _ .
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| The NRC inspectors reviewed Colt SIL, Issue 7 (December 16,1985), entitled
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| "Intercooler Spacer Bar," to determine if information therein pertained to the Wolf Creek EDGs. This SIL, which had not been received by the licensee, .
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| addressed a potential problem associated.with the spacer bar supporting the side-mounted turbocharger intercoolers and noted that the spacer bar mounting bolts should be periodically checked for tightnes The NRC inspectors performed a field walkdown of the A and B EDG intercooler supports and associated cooling water piping. During the walkdown, the inspec-tors noted that all four turbocharger cooling water piping lines were missing a seismic and vibration control pipe support that was required to be installed by vendor's design drawings. In response to this observation, the licensee con-tacted the vendor (Colt Industries) to determine if the turbocharger cooling pipe could perform its intended function without the seismic and vibration control supports and whether the turbocharger cooling pipe would experience cracking or the flange bolts would loosen as a result of excessive vibratio The vendor referred the licensee to Colt Industries' Engineering Report N M-018-0367-02, "Seismic Calculations.for Skid Mounted Piping." A table in this report indicated that the support bracket would be required for the turbo-charger cooling piping in a seismic event if the length of the piping was greater than 60.7 inches. The licensee measured the subject piping and found that it was 56 inches in length; thus concluding that the turbocharger cooling pipe could perform its intended function during a seismic event without the support bracket. The licensee gave the NRC inspectors a draft copy of their engineering seismic calculation, which also indicated that the pipe did not require the support to withstand seismic loadin Even though the available engineering data did not support installation of the supports for seismic reasons, Colt urged the licensee to install the four missing supports to ensure that vibrations from the operating diesel engine would not cause degradation of engine components. In addition, Colt recommended that the licensee visually inspect the pipes for cracking and a loss of jacket cooling water and perform a torque inspection for all associated pipe flange bolt The licensee took imediate actions to fabricate and install the four pipe supports and performed the inspections recommended by Col During those inspections, quality control inspectors found that the turbocharger cooling piping on the A EDG contained a weld defect. This item was referred to engineering for further evaluation. In addition, when it was determined the two of the flange bolts were torqued below minimum requirements, a work request was issued to retorque all of the affected pipe flange bolts for both emergency diesel engines. Before the conclusion of the inspection, the
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| licensee further comitted to perform nondestructive examinations on all four turbocharger cooling water pipes and a vibration test and analysis to determine if there were any additional adverse effects on the cooling pipe caused by operating the EDGs without the support At the time the EDGs were originally constructed at Wolf Creek, the turbocharger cooling water piping vibration supports were not installed as required by the vendor's design drawing. During the installation work, licensee personnel who were res)onsible for verifying that the EDGs were properly constructed did
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| , not ensure t1at the supports had been installed. This failure to verify that the as-built configuration of tne EDGs was consistent with the Colt design
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| drawing is a potential enforpement finding (Item No. 88-200-5).
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| 3.5.2 Results Sumary The NRC inspectors identified several instances where the licensee did not -
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| obtain and evaluate all the applicable EDG vendor information (SILs). In part, this contributed to the four missing pipe supports for the cooling water piping lines not being discovered by the licensee. Although the supports were not necessary for seismic support, the vendor did recommend that they be added for vibration reasons. QC inspections of the EDGs during this inspection did reveal that excessive EDG operational vibrations had caused the pipe flange bolts to loosen to the point wht:re they did not meet torque requirement These issues point to the need for additional attention to detail in the area of vendor interfac . Diesel Generator Jacket Water Pressure Transmitters 3. Inspection Results The NRC inspectors performed a walkdown of the A and B EDGs and their associated support systems. During the walkdown, the NRC inspectors noticed plant modifi-cation reauest (PMR) Tag No. 20315, dated April 11, 1986, adjacent to the jacket water pressure indicator gauge on the local control panel for the A ED The information on the tag indicated that a pulse in the gauge's serising line was causing a false indication on the pressure gauge. The NRC inspectors went to the local control panel for the B EDG to determine if the same con-dition existed. They saw two information tags located next to pressure gauges for the jacket cooling water and the jacket water intercooler. Both informa-tion tags indicated that there was a sensing line pulsation problem and that the lines were valved out to isolate the system and stop spurious alarms in the control room during system testing. The NRC inspectors asked a plant operator if it was possible that the sensing line for the indicator gauge on the A EDG was isolated even though there was no indication of such on the PMR tag. The operator stated that the PMR tag did not serve that purpose and that the line for the A EDG should not be isolated. However, when the NRC inspectors and the operator examined the line, they found it had been valved out and isolated. In response, the operator notified the SR0 on duty and replaced the PMR tag with one containing the correct line configuration information. Because the licensee took immediate action to correct the problem and because the line was used for indication of system operating parameters, the inspectors have con-sidered this issue adequately resolve The NRC inspectors interviewed cognizant instrumentation and control (I&C)
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| personnel to determine why the false indication conditions existed and what had been done to correct the problem. Previously, a temporary modification was implemented to install pressure damping devices in the sensing line. The dempers alleviated the problem until they became clogged with impurities from the jacket cooling water. Subsequentl The licensee then initiated EER 87-KJ-01 (y, the dampers were removed. June 9, 1987) to The EER contained information indicating that the problems with the pressure transmitters resulted from pressure pulsations in the jacket water sensing line side of the transmitter. In the disposition of the EER, Plant Engineering recommended remounting the pressure transmitters and placing dampers or similar flow restriction devices adjacent to the transmitter where the trans-mitter sensing line ties into the pressure portion of the system. This i
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| * ' | | Sincerely, Odginal Signed By n A. B. Denh L. J. Callan, Director o@8 Division of Reactor Projects $6 Enclosure: |
| | g, Appendix - Notice of Violation i No 38 cc w/ enclosure: (see next page) ! |
| | *RIV:RI *C:0PS C:PS *D:DRS D:DRP C 53 DRHunter/ld JEGagliardo TStetka JLHilhoan LJCallan "Of ho / /88 / /88 (o/f/88 / /88 W/ti/88 |
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| modification, when implemented, will shorten the length of line between the transmitter and damper and reduce the amount of impurities that could clog the dampers. I&C personnel stated that, although this modification is planned, the problem is still ongoing. The licensee has not to date considered the cause of the pulsations ar.d the effect of the proposed corrective actions. This failure to aggressively pursue the cause and take action to stop the sensing line pulsations that have existed since 1986 is considered a potential enforce-mentfinding(ItemNo. 88-200-1f).
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| 3.6.2 Results Summery The NRC inspectors determined that the licensee had not adequately addressed the malfunctions in the jacket water pressure sensing line and instruments of the EDGs. Since initial discovery of the problem in April 1986 to the time of the QVFI, the licensee has not aggressively pursued the cause o.f the pulsations in the system nor have they implemented timely, effective corrective actions to ensure accurate and reliable system performance. Disregard of this instrument's inability to perform its intended function is not an attribute of prudent, safe operation of the EDG syste INDEPENDENT SAFETY REVIEW ORGANIZATIONS The NRC inspectors reviewed the activities of Wolf Creek's independent safety review groups to determine their effectiveness and contribution to the plant's safe and reliable operatio .1 Pla It Safety Review Comittee (PSRC)
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| 4. Inspection Results The NRC inspators reviewed the minutes of six PSRC meetings (306, 316, 317, 319, 320, and 322), interviewed selected personnel with regard to the PSRC activities, and attended a PSRC meeting (No. 322) on June 14, 198 The PSRC function is specified by Procedure ADM 01-002, Revision 16, "Plant Safety Review Committee." The procedure implements the requirements of Tech-nical Specification 6.5.1, "Plant S. fety Review Comittee (PSRC)." The PSRC reetings were conducted routinely at weekly intervals, which is more frequently than required by the Technical Specifications. Additional meetings-were scheduled when deemed appropriat The QA manager, or a designated alternate, normally attends the scheduled PSRC meetings, even though the QA manager is not a membe The NRC inspectors determined that all but two of the selected PSRC members had the experience and equivalent training normally required to take an exami-nation for a senior reactor operator's license at Wolf Creek. The two PSRC members with less exte.nsive training were the Manager of Maintenance and Modifications and the Manager of Plant Support. The inspectors discussed upgrading the training of these two managers with the license The NRC inspectors reviewed the materials discussed during the PSRC meeting (322) conducted on June 14, 1988. During the meeting, plant modification request PMR 02577, Revision 0, "Penetration Roundary Change," was reviewed, l The PMR had been processed in response to corrective work request (VR 00688-88)
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| dated February 9,1988. The WR was written to ducument that the top 6 inches
| | Wolf Creek Nuclear Operating ~ -2-Corporation Wolf-Creek Nuclear Operating Corporation ATTN: Otto Maynard, Manager of Licensing P.O. Box 411 Burlington, Kansas 66839 Wolf Creek Nuclear Operating Corporation ATTN: Gary Boyer Plant Manager P.O. Box 411 Burlington, Kansas 66839 l Kansas Corporation Commission ATTN: Robert D. Elliott, Chief Engineer Fourth Floor. Docking State Office Building Topeka, Kansas 66612-1571 |
| | | . Kansas Radiation Control Program Director bectoDMB(IE01) |
| | l bec distrib. by RIV: |
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| | l RRI R. D. Martin, RA l SectionChief(DRP/A) DRP RPB-DRSS B. DeFayette, RIII l |
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| | RIY File SRI, Callaway, RIII MIS System RSTS Operator l Project Engineer. DRP/A Lisa Shea, RM/ALF P. O'Connor, NRR Project Manager DRS Gary M. Holahan, NRR R. P. Correia. NRR F. C. Hawkins, NRR S. H. Weiss, NRR J. Gagliardo D. Hunter G. Sanborn l |
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| of Radflex material was missing from penetration OP 142S1099 located on eleva-tion 2026' of the auxiliary building. The shift supervisor declared the p netration operable on February 9, 1988. The initial review of the degrace condition of the sealant was completed on February 18, 1988, and resulted in a "use-as-is" disposition of the WR. The basis for the use-as-is disposition was that there was enough Radflex material remaining to allow sufficient fire rating but not enough for a radiation barrier. As a result, the design of the penetration seal was revised from an RB-9 type (Radflex) to an M-9 (fire seal).
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| The followup engineering disposition regarding the condition of the penetra-tion was completed on May 3, 1988, and concluded that the floor at elevation 2026' separates fire area boundaries and requires a 3-hour fire-rated penetra-tion seal. Because of the uncertainty of the current consistency of the Radflex material in penetration OP 142S1099, engineering could not establish that the penetration would meet these fire qualification testing requirement WCGS's Updated Safety Analysis Report (USAR), Section 9.5, Table 9.5.1-3, requires that all fire barriers and their penetrations separating safety-related areas from those that are not safety related or separating portfans of redundant systems important'to safe shutdown shall be operable at all time Should one or more be found to be inoperable, a continuous fire watch on one side of the affected barrier or an hourly fire watch patrol must be estab-lished within 1 hour. The inspectors discussed the May 3 engineering evalua-tion and the degraded condition of the fire seal with the licensee. On June 14, 1988, the licensee issued Fire Protection Impairment Control Permit No.88-244 to establish a firewatch. In effect, a fire watch should have been established within 1 hour from the time the fire seal was determined not to meet fire qualification testing requirements. This failure to implement the required fire watch between May 3 and June 14, 1988, is considered a potential enforce-ment finding (Iter No. 88-200-6).
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| 4.1.2 Results Summary The NRC inspectors determined that, with the exception that a required fire:
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| watch for an unqualified penetration fire barrier was not established, the PSRC function was established and functioning as required by Technical Spucifica-tion .2 Nuclear Safety Review Committee (NSRC)
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| 4. Inspection Results The NRC inspectors reviewed the minutes of NSRC meetings conducted in 1987 and 1988 and interviewed selected personnel with regard tr NSRC activitie The NSRC function is specified by Policy No. II.13.0, Revision 3, "Nuclear Smfety Review Committee Charter." The policy implements the requirements of Technical Specification 6.5.2. Document reviews and discussions revealed that the meetings were scheduled and conducted more frequently than required--
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| generally three or four times per year. NSRC meetings are routinels conducted at the site training center and include a scheduled plant tour. Also, the members can independently review specific areas of plant operations, such as operations, chemistry, and health physics. The requirements of NSRC audits is
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| addressed in detail, including overall responsibility, planning and implemen-tation, audit reports, and resolution of findings. The audits and audit results are maintained in an action item list, as reflected in the NSRC meetirig minute l
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| 4.2.2 Results Sumary The NRC ins]ectors determined that the NSRC consisted of technically capable personnel w1o fulfill the requirements of the Technical Specifications. The NSRC has provided upper management with technically sound recomendations concerning plant safety and reliability and are functioning as an effective quality verification organizatio '
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| 4.3 Nuclear Swfety Engineering (NSE)
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| 4. Inspection Results The NRC inspectors reviewed selected NSE reviews and evaluations to determine the effectiveness of NSE as an independent quality verification organizatio The NSE function is specified by Procedure KP-750, Revision 0, "Statement of Responsibilities Nuclear Safety Engineering." The procedure implements Item I.B.1.2 of NRC N, REG-0737 Technical Specification 6.2.3, USAR Chapter 18.1.7.2, and outlines actions in response to NRC Generic Letter 83-028. NSE performs surveillances of plant activities in accordance with the requirements of Procedure KP-751, Revision 0, "Surveillance of WCGS Activities by Nuclear Safety Engineering." The procedure provided definition, responsibilitiss, and the scope of'the surveillance activitier for NS The NSE also reviews almost all operational information concerning other comercial nuclear power facilitie It routinely receives all reactor trip data and is required to complete the independent review of all unscheduled reactor trips before reactor restart if the trip was complicated by other plant perturbation Recently, the NSRC requested NSE to investigate a 4-percent indicated decrease in total reactor coolant system,(RCS) flow. NSE determined that an analysis of the calibration data for the RCS narrow-range resistance temperature detec-tors (RTDs), which were used to establish core enthalpy rise, was required because an increase of 1.5 to 2.0'F had been identified. The review of the RTD calibration data taken during the 1987 outage was compared to the data taken during initial startup in 1985. The comparison indicated (1) a much wider variation between the hot leg RTDs (but not exhibited between the cold leg RTDs) and (2) a disparity between the hot leg and cold leg RTDs. The wider variation exhibited by the hot leg RTDs and the disparity between the hot and cold leg RTDs indicated that the hot leg RTDs output signals had drifted differently ture gradients than(and the resultant cold leg RTDs, thermalpossibly)as a result of stress experienced bythethesteep tempera-hot leg RTDs following a reactor trip (a large number of which occurred during the first and second year of plant operation).
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| The licensee had implemented a number of actions to attempt to reduce the RTD errors, including the Westinghouse error analysis methodology. These actions resulted in a reduction in the RTD errors and an increase in the indicated (calculated) RCS flow. However, the NRC inspectors were concerned that the
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| routine use of the Westinghouse error analysis methodology (cross-calibration of RTDs and development of correction factors) and the utilization of RTD vendor supplied resistance (R) versus temperature (T) curves may not be con- .
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| servative, in that the RTDs at Wolf Creek (or any other ft cility which uses such methodology) may never be calibrated to a known standard to ensure generic senser drift does not occur during the 40-year lifetime of the plant. Wolf .
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| Creek does not use the RTDs installed in thermowells to calibrate, under !
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| controlled conditions, the RTDs in the protection system (imersion-type RTDs).
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| These system RTDs have not been checked to a known standard, directly or indirectly, since initial installatio Items 7 and 8 ("Overtemperature Delta T" and "Overpower Delta T," respectively)
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| in Technical Specification 3/4.3.1, "Reactor Trip System Instrumentation,"
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| Table 4.3-1, specify that a channnel calibration is to be performed at least once every 18 months. Technical Specification 1.5, "Channel Calibration,"
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| specifies in part that a channel calibration shall be adjusted, as necessary, such that the channel responds within the required range and accuracy to known values of input and shall encompass the entire channel including the sensor The methodology used to calibrate the RTDs does not include checking the accuracy of the RTDs to known values of input (temperaturo). Shifts in the RTD calibration curveRTD out-of-tolerance mayoutput not bevalues detected in a timely (Observation manner, Item No. which ma88-200-7)y
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| . The NRC result in inspectors discussed this matter with the licensee; it vill require further NRC NRR staff revie .3.2 Results Sumary The NRC inspectors determined that the NSE appeared to be an effective, tech-nically-oriented organization. The NSE has provided management with extensive and accurate assessments of plant issues, such as the RTD cross calibration issue and the problems with the control room chlorine monitor INDUSTRY TECHNICAL INFORMATION PROGRA:4 (ITIP)
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| 1 Inspection Results The ITIP function is specified by Procedure KGP-1311, Revision 1, "Industry Technical Information Program." The ITIP implements the licensee's response to items addressed in NRC NUREG-0737, Item I.C.5, "Procedures for Feedback of Operating Experience to Plant Staff."
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| The NRC inspectors reviewed evaluations of twelve ITIP items received by the licensee, as well as selected monthly status reports, a recent QA audit report, the most recent effectiveness review report, and Procedure KGP-1311. The inspectors held discussions with selected licensee personnel wi6 h regard to ITIP activitie The NRC inspectors' review of the completed ITIP evaluations indicated that the timeliness of the reviews had improved dramatically over the past 3 month The timeliness issue was previously identified in QA Audit Report TE:50140-K202, dated March 23, 198 The report specifically identified the lack of timeli-ness of the initial evaluations, a significant backlog of items requiring reviews, and the need to complete programatic changes expeditiousl The
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| evaluation review times have recently decreased from months to days. Discuss-ions revealed that che licensee was applying additional effort to decrease the ITIP backlog and other programatic improvements have been complete .2 Results Sumary The NRC ins)ectors determined that weaknesses noted by the licensee's QA organization regarding t1e timeliness of reviews have recently improved. However, the importance of evaluating industry information on plant equipment and components in a timely way is necessary for reliable and safe operations. Section 3.5 of this report provides details of the ramifications when the ITIP fails to fulfill its required function. Other ITIP functions were implemented in accordance with applicable WCGS procedure ACTIVITY / EVENT REVIEW The NRC inspectors reviewed the effectiveness of the licensee's quality verifi-cation organizations through the corrective actions associated with four specific activities: (1) emergency service water pipe wall thinning, (2)
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| ra pressurizer leakage, and sp(4)y valve replacement containment packing box, (3) reactor vessel head 0-ring cooler A repai .1 Emergency Service Water (ESW) Pipe Wall Thinning 6. Inspection Results The NRC inspectors reviewed documents and interviewed licensee personnel with regard to pipe wall thinning experienced in portions of the ESW system in 1985 during normal system operations. Pipe wall thinning appeared to be caused by erosion / corrosion from combinations of elevated flow rates through throttled butterfly valves and the configuration of the ESW syste With the exception of several short outages resulting from equipment malfunc-tions, the unit operated continuously until the comencement of the refueling outage in September 1987. The NRC inspectors reviewed a number cf specific activities related to the corrective actions associated with pipe wall thinnin WorkRequest(WR) 00653-87 was issued on February 13, 1987, fact that the ESW piping below valve EFV-058 (throttled butterfly valvedocumenting)the was less than the specified minimum pipe wall thickness of 0.328 inches in numerous locations. The WR noted that the system was operable and the condition not reportable per 10 CFR 50.72. The WR was forwarded to Nuclear Plant Engineering (NPE for evaluation and an engineering disposition was provided on February 19, 1987, specifying that repair of the minimum wall for pipe spool piece 1-EF05-S-005/142 should be re) aired per instructions in Plant Modification Request (PMR) 1903. PMR 1903 1ad been used to repair train "B" of the ESW system during the 1986 refueling outage. The weld overlay repair of the ESW piping was subsequently performed during June 26 to July 1, 198 The required system leak test was performed on July 1,198 , the Nuclear Safety Engineering (NSE) group performed a surveil-In May(SSR lance 87-045) of selected activities associated with the ESW pipe minimum wall thickness deficiency. A draft report of SSR 87-04S was provided to the plant manager on May 14, 1987. In the draft, NSE noted that no justification for continued operation had been provided regarding the thin-wall ESW system piping in that the engineering evaluation request (EER) only addressed the
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| final weld overlay repair condition. The NSRC chairman, made aware of the issue by the NSE, also pursued the questionable condition of train A of the ESW system. Independent calculations were also performed by the licensee's .
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| engineering staff that confirmed that the ESW did not meet all its design requirements. Wolf Creek Updated Safety Analysis Report 0 SAR) S.ction 9.2.1.2.1.1 states that the ESW piping and valves are designed to the require-ments of ASME Section III, Class 3. Section 9.2.1.2.1.1 of the USAR states that the ESW is safety-related, is required to function following a Design Basis Earthquake (DBA), and is required to achieve and rLaintain the plant in a safe shutdown conditio The report of SSR 87-045, dated June 4, 1987, identified three concerns regard-ing the handling of minimum wall work requests, including (1) the operability determination made by the shift supervisor, (2) availability of information to operations, and (3) a defined program for handling pipe erosion. Their report also stated that the current safety evaluation covers only the permanent repair and not the justification for the continued operability of the component during the interim perio The NSRC, aware of the EWS wall thinning issue in May 1987 as a result of NSE involvement, held discussions with plant management, including the Chief Executive Officer (CE0), on June 2,1987. The NSRC Chairman was intimately involved in the oral and written communications regardin I
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| by the NSE surveillance (SSR 87-045, dated June 4,performed 1987)g the concerns raised in mid-May 1 1987. On June 18 1987, the NSRC chairman established a special review group (two consultants),to evaluate the ESW wall thinning matter. The issue was discussed in detail in NSRC meeting 87-02, conducted on June 24 and 25,1987, and specific recommendations were proviue; to the licensee's CEO by letter on July 10, 1987 for consideration. The recommendations included the following:
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| (1) Further encourage and formalize the communications process between project personnel in the area of nonconformance report (NCR) disposition and implementatio (2) Require a study and documentation of the lessons learned as a result of this occurrence with ESW pipe wall thinning from February 1987 until the repair is completed. Additionally, the CEO was provided a copy of the NSE surveillance report of SSR 87-045, which could not be closed because corrective actions have not been complete The licensee's CEO provided a draft letter on the subject to the three vice presidents (engineering, quality, and nuclear operations) on July 28, 198 The draft letter addressed three matters regarding the "detert:Ination of operability," including (1) the operations group responsibilities, (2) NPE evaluations of nonconforming conditions, and (3) provision of the NPE evalu-a+. ion to operations for reassessmen Subsequently, plant Administrative Procedure ADM 08-212 and Engineering Procedure KPN-314 were revised to provide an erosion / corrosion program and here the system design provide functionnotification is adverselyofaffecte operations Operations by engineering Order w(0P) 87-110 was issued on July 29, 1987, to instruct that all thin-wall piping problems would be docu-mented by an NCR in accordance with the procedure. To enhance connunications betwcen operations and engineering, engineering now attends work planning reeting Licensee personnel stated that they will, in the future, request 19 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
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| , i-l temporary relief from ASME Section XI requirements from the NRC, when required, '
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| and provide technical justification for continued operation for each relief request. NPE had responded to the NSE surveillance report (SSR 87-045) on .
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| February 2, 1988. Nuclear Operations had not responded at the time of this inspectio The NRC inspectors determined that the licensee's actions were inappropriate immediately following the discovery of the degraded ESW piping on February 13, 1987. The engineering disposition completed on February 19, 1987, stated that the degraded ESW pipe downstream of valve EFV 058 required repairs because it did not meet ASME requirements and, therefore, may not be capable of performing its specified safety functions. At this time, the licensee did not declare the system inoperable and allowed it to remain inservice until repairs were begun on June 26, 1987. Wolf Creek Technical Specification Limiting Condition for Operation (LCO) 3.7.4 requires at least two independent ESW loops be operabl In addition, with only one ESW loop operable, the inoperable ESW loop must be restored to operable status within 72 hours or the reactor must be in at least hot standby within 6 hours and in cold shutdown within the following 30 hour The licensee's failure to declare the ESW system inoperable and meet the re'quirements of LC0 3.7.4 is considered a potential enforcement finding (Item No. 88-200-8). This issue was previously discussed in Region IV Inspection Report 87-15, dated July 22, 198 .1.2 Results Sumary The lack of timeliness and thoroughness of the operational response to the ESW pipe wall thinning and the apparent lack of coordination between the plant operations staff and the various technical support groups is considered a significant weakness in the licensee's operability determination proces This recurrance of a problem with ESW pipe wall thinning should have triggered an immediate response by the licensee to declare the system inoperable and procede with expeditious repairs to assure safe and reliable system operatio .2 Pressurizer Spray Valve Replacement Packing 3ox '
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| 6. Ir.spection Results The NRC inspectors reviewed the documented work activities associated with the replacement of the pressurizer spray valve packing box. The work was completed on December 29, 1987, in accordance with work request WR 00101-87, dated January 1,1987, and ASME Section XI Plan No. RR-87-074, dated September 21, 198 The NRC inspectors reviewed documents and interviewed licensee personnel with regard to the work completion review by the licensee for the pressurizer spray valve packing box replacement on March 18, 1988. The ap3ropriate component qualific& tion documentation, which was supposed to have )een part of the pro-curement package from Wes inghouse, had not been received. As a result, the licensee could not ensure that the assembly of the pressurizer spray valve packing box met ASME Section XI requirements. The missing documents were the required ASME Code Data Report, Certified Material Test Reports (CMTRs), and nondestructive examination (NDEP reports. Even so, the plant was restarted in late December 1987 with the ASME pressure boundary component of undetermired quality installed in the reactor coolant syste Engineering analyses did wt determine that the the component was acceptable and ASME Section XI relief .ws
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| *
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| not.obtained until late May 1988 (WR 1285-88, dated March 17, 1988 completed on May 24, 1988; and PMR 02535, dated April 25,1988).
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| The licensee initiated a Programmatic Deficiency Report (PDR MM 88-07, dated April 18, 1988) documenting a deficiency with the pressurizer spray valve pack-ing box. The Manager of Purchasing and Material Services was given the assign-ment to coordinate and collect documentation of the completed actions in response to the PDR. The PDR addressed the WCGS program and Westinghouse procedure requirements and the immediate corrective actions to investigate the apparent Westinghouse error in omitting the appropriate component qualification documentation. In addition, a review of all ASME Code Section III documenta-tion packages that had been received prior to installing an ASME Section III boundary item was to be performed and the spare spray valve packir.g box assembly procured at the same time as the installed component was to be rejected. The cause of the issue was to be determined and corrective actions to prevent its recurrence were to be performed. The NRC inspectors determined that all but two of the actions were completed. The rescheduled completion date for the open items was September 1, 198 The overall corrective actions included the review of other repair-replacement work packages to ensure that no other nonconforming safety-related equipment had been installed. This review revealed that 17 issues required comment resolution. However, none of these issues were deemed by the licensee to impact safety-related equipment operabilit The licensee plans to replace the nonconforming valve packing box assembly during the next outage of sufficient duration and proper plant conditions to permit replacement. The NRC inspectors concluded that the licensee's overall corrective actions for this issue appeared to be extensive and acceptabl .2.2 Results Summary .
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| The licensee's failure to ensure that the spray valve packing box conformed to ASME requirements before plant startup has been addressed in NRC Inspection-Report 50-482/88-15. The licensee's actions taken once the deficiency was identified were both timely and effective and should ensure that repetition will not occu .3 Reactor Vessel Head 0-Ring Leakage
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| 6. Inspection Results The NRC inspectors reviewed the documented activities associated with the reactor vessel head 0-ring leakage event of December 26, 1987, through Janaury 21, 198 Leakage from the inner reactor vessel (RV) head 0-ring occurred on December 26, 1987, during plant startup after the scheduled refueling outage. The inner 0-ring leakoff path was isolated and the leakage system aligned to monitor the outer 0-ring. Subsequently, additional leakage was detected when the leakoff temperature from the outer 0-ring increased between January 19 and 21, 198 The reactor was cooled down and the RV head removed on January 26, 198 Between January 26 and February 2,1988, the RV head 0-ring seating area was inspected and cleaned. The licencee determined that the RCS level was at too
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| high a level during the previous RV head placement, allowing water to overflow into the 0-ring channel and eventually leak into the leakage indication syste On February 2,1988, the licensee commenced installation of the new 0-rings ,
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| with the RCS cool nt level between 12 and 39 inches below the RV flange leve The RV head wat at and the RV studs imediately installed and torqued while the RCS water letel was held at 39 inches below the RV flange. Subsequent RV head 0-ring performance has been satisfactor The NRC inspectors determined by document review and interviews that corrective actions associated with this event were extensive and included specific program and procedure improvements. Further, the installation of a new RV water level indication system during the next refueling outage was scheduled and should improve the control of the RCS water level during outage .3.2 Results Summary The licensee's inattention to detail during the installation of the RV head 0-rings during the 1987 refueling outage resulted in a forced plant shutdown as a result of leakage of reactor coolant past the 0-rings. The licensee's maintenance staff did not provide quality workmanship and the quality verifi-cation personnel involved did not identify the deficiencies, which would have prevented the forced shutdown. (Reference NRC Inspection Report 50-482/88-04)
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| In addition, the removal and replacement of RV head 0-rings resulted in many additional staff-rem exposure hours, which is a concern to the ALARA commitment (as low as reasonably achievable). The additional stress placed on plant personnel as well as systems and components as a result of the shutdown and followup activities also is significant from a safety standpoin The NRC inspectors determined that the overall corrective actions taken by the licensee regarding this matter appeared to be acceptable. These actions should provide improved cleanliness controls associated with the RV head, adequately inspected RV 0-rings, and an enhanced RCS water level indication system. The RCS new water level indication system also will provide better control of the RCS water leve ,
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| 6.4 Containment Cooler A Repair 6. Inspection Results The NRC inspectors reviewed documented work activity associated with the repair of the A containment cooler to assess the work planning and the quality assurance / quality control (QA/QC) involvement and conducted selected interviews to provide an adequate understanding of the work plan and associated activitie The A containment cooler leak was identified and a work request (WR 04105-87)
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| initiated on October 165 1987. The licensee conducted a substantial amount of planning, repair selections, engineering, repair work, and testing during October 20 to December 22, 1987. The repaired cooler was tested on December 1, 1987, the cooler support reinspected on December 14, 1987, and the cooler was returned to service on December 15, 1987. The insulation was replaced on December 18, 1987. The final maintenance review was performed by December 22, 1987, and the final quality review was complete on February 7,198 Subsequent review by the licensee in April 1988 of the completed work package revealed two deficiencies: (1) no certified material test report from a
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| qualified vendor was available for the brazing material used to repair the coolercoilsand(2)repairstothetubesoftheheatexchangercoilwereper-formed without a qualified brazing procedure and a procedure qualificatio record for the base metal and thickness required. The specific deficiencies were brought to the attention of engineering and the subsequent engineering dispositions concluded that the items were acceptable. Programmatic deficiency reports were v.-itten on May 26, 1988 (PDR OP 88-095, improper brazing material), and April 29, 1988 (PDR OP 88-094, improper brazing procedure),
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| specifying corrective actions wt h completion verified as of May 27, 198 The corrective actions regarding the issuing of improper brazing material appeared to be satisfactory. However, the NRC inspectors noted that the weld data sheet (ADM08-300, Exhibit A) was initialed and dated by the weld engineer, without realizing that the incorrect brazing procedure was specified. The weld engineer routinely uses a desk-top procedure and checklist that are not part of the applicable administrative procedure. The NRC inspectors discussed the use of uncontrolled desk-top procedures and simple checklists with licensee personne Additionally, the quality control review required to signify agreement with specified brazing requirements was initialed and dated on the weld data sheet withcut ensuring the correct brazing procedure was specified. After the inspectors discussed.this matter with the licensee, an additional PDR was written on June 14, 1988 (PDR QC-88-011, QPS review of WR04105-87 failed to identify the wrong brazing procedure was to be used in the field). The PDR noted that a memorandum was written to all QC quality plant support personnel reminding them what type of review is required by Procedure QP 12.1, paragraph 7.1.2, and that QF 12.1 is to be revised to enhance review process to clearly state what documen,ts are to be reviewe .4.2 Results Summary The NRC inspectors determined that the corrective actions taken by the licensee regarding the cooler repair appeared to be adequate. However, the NRC inspec-tors concluded that the final maintenance and QC review to determine technical adequacy and completeness of the package should have been completed before the shift supervisor restored the system to servic . EXIT INTETJIEW The NRC inspectors held meetings with licensee supervisory and management per-sonnel periodically during the course of the inspection to discuss the status of tne inspection. The NRC inspectors met with the licensee's representatives
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| . (19cluded in the list in Appendix A to this report) on June 17, 1988, to summar-ize the inspection scope and findings end the to discuss the observations and potential enforcement findings. Although proprietary material was reviewed during the inspection, no proprietary material is contained in this repor ._
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| , ,e 6 .
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| APPENDIX A
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| PERSONS CONTACTED Wolf Creek Nuclear Operatir$g Corporation Personnel
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| *B. D. Withers, President
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| *F. T. Rhodes, Vice President Nuclear Operations | |
| *R. M. Grant, Vice President Quality
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| *J. A. Bailey, Vice President Engineering and Technical Services
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| *C. D. Boyer, Plant Manager
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| *C. E. Parry, Manager-Quality Assurance
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| *G. W. Reeves, Manager-Quality Contr?1
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| *W. M. Lindsay, Manager-Quality Eva'.sation
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| *R. H. Belote, Manager-Nuclear Safety Engineering
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| *J. M. Pippin, Manager-Nuclear Plant Engineering | |
| *A. A. Freitag, Manager-Nuclear Plant Engineer- *R. W. Holloway, Manager-Maintenance and Modification
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| *M. G. Williams, Manager-Plant Support | |
| *0. L. Maynard, Manager-Licensing
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| *S. Wideman, Licensing
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| *K. Peterson, Supervisor-Licensing
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| *J. A. Zell, Manager-Training
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| *C. W. Fowler, Manager-ISC
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| *R. J. Potter, Manager-Material / Supplier Quality
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| *W. B. Wood, General Counsel
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| *J. L. Houghton, Supervisor Operations
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| *M. L. Johnson, Nuclear Coordinator-KG&E
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| * B. Norton, Supervisor Reactor Engineering
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| *L. Payne, Supervisor Quality Plant Support
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| *R. E. Gimple, Supervisor Materials Quality
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| *C. G. Patrick, Supervisor Quality Systems
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| *C. J. Hoch, Quality Assurance Technician R. S. Benedict, Manager Plant Inspection R. S. Robinson, Supervisor, I&C Maintenance W. G. Eales, Jr., Manager Electrical Systems Engineering N. Hoadley, Lead Engineer, Nuclear Plant Engineering A. Clason, Manager Engineering Support T. Deddens. Outage Manager L. Stevens, Lead Engineer Nuclear Plant Engineering Other licensee employees contacted included operators, engineers, auditors, technicians, mechanics, and office personne NRC Personnel
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| *F. C. Hawkins, Chief, Quality Operations Section, NRR
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| *P. W. O'Connor, Project Manager, NRR
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| *J. Jaudon, Deputy Director, Division of Reactor Safety, RIV
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| *B. Bartlett, Wolf Creek Senior Resident Inspector, RIV M. E. Skow, Wolf Creek Resident Inspector, RIV
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| *B. Little, Callaway Senior Resident Inspector, RIII
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| * Denote those attending the exit meeting on [[Exit meeting date::June 17, 1988]]
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| }} | | }} |
