IR 05000456/1999005
| ML20207A595 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/20/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20207A592 | List: |
| References | |
| 50-456-99-05, 50-456-99-5, 50-457-99-05, 50-457-99-5, NUDOCS 9905270066 | |
| Download: ML20207A595 (36) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION lli l
Docket Nos: 50-456, 50-457 License Nos: NPF-72, NPF-77 l
l Report No: 50-456/99005(DRS); 50-457/99005(DRS)
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- Licensee
- Commonwealth Edison Company l
l Facility: Braidwood Nuclear Plant, Units 1 and 2
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Location: RR #1, Box 84 Braceville, IL 60407
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Dates: April 5 through April 23,1999
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hspectors: V. P. Lougheed, Team Leader l C. H. Brown, Reactor inspector 1. N. Jackiw, Reactor inspector D. E. Jones, Reactor inspector l G. F. O'Dwyer, Reactor inspector l D. S. Schrum, Reactor inspector i H. S. Anderson, Contractor Approved by: John M. Jacobson, Chief Mechanical Engineering Branch !
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9905270066 990520 PDR ADOCK 050004 6 G
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EXECUTIVE SUMMARY Braidwood NucMar Plant, Units 1 and 2 NRC Inspection Report 50-456/99005(DRS); 50-457/99005(DRS)
l This was an announced team inspechon of approximately three weeks duration to review engineering and technical support and corrective actions. The inspection included a review of l
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50.59 evaluations and screenings as well as a review of selected NRC issues from previous inspections. The following statements summarize the inspection results:
- Based on the inspection results, the inspectors concluded that the methods used to control design changes and modifications at the Braidwood plant were effective. The modifications were adequately designed and of good technical quality. Appropriate controls were established over installation activities. Post modification testing was very good, especially in the case of the Woodward govemor replacement. (Section E1.1)
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The 10 CFR 50.59 evaluations and screenings were effective, thorough and appropriate to the plant changes. (Section E1.2)-
- Temporary modifications were well controlled, including having stringent controls to ensure that temporary modification were removed when no longer neede (Section E1.3)
- While the final numerical results were acceptable, calculations, at times, contained unverified assumptions or used incorrect methodologies. There was good correlation I among the calculations, the technical specifications, and the updated final safety analysis l
report. (Section E1.4)
.' Inservice testing surveillances for the residual heat removal and safety injection pumps ensured that both the American Society of Mechanical Engineers (ASME) Code requirements and the design basis requirements were satisfied. (Section E1.5)
- Adequate engineering interface and support was provided to the modification proces Final closeout of the modifications occurred in a timely fashion. (Section E2.1)
- The operability determination process was effective. Adequate bases were provided for the operability determinations and the evaluations could be followed without difficult (Section E2.2)
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. The methods used to obtain and disposition industry operating experience were !
effective. A strength noted was the support provided to the Operating Experience Program Coordinator by the corporate and other site coordinators. (Section E2.3) j
- The emergency lighting units maintenance rule performance criteria of 15 functional failures per month was not adequate and could have resulted in masking performance !
problems with emergency lighting units. A non-cited violation was issued as immediate l corrective actions were taken and the issue was entered into the corrective action system. (Section E2.4)
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The licensee failed to identify potentially adverse battery conditions due to an inadequate surveillance procedura and inadequate surveillance procedure reviews. A non-cited violation was issued as immediate corrective actions were taken and the issue was entered into the corrective action system. (Section E2.5)
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The verification and the tracking of engineering training appeared satisfactory. The program, procedures and certification guides provided a sufficient level of trainin (Section E5.1)
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The onsite and offsite review committees were effective in performing their assigned I
reviews, investigations and evaluations. Members of these committees were aggressive in pursuing plant problems and issues. (Section E6.2)
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Actions taken once a problem was identified were indicative of a strong corrective action program. The rmlority of plant problems appeared to be identified, assessed, and had corrective actioe assigned. (Sechon E7.1)
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Root cause investigations were accurate and thorough. The licensee's trending program and effectiveness reviews contributed to identifying repetitive problems which required additional corrective actions. (Section E7.2)
The assessment process used by the licensee's independent oversight group met the requirements of 10 CFR Part 50, Appendix B. The continuous assessment process provided an acceptable review of the area being assessed. The self-assessment program provided sufficiently in-depth information so that the assessed group could make the appropriate corrections in the areas of weakness. (Section E7.3)
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Report Details 111. Enaineerina E1 Conduct of Engir,eering E1.1 Desian Chances and Modifications Insoection Scope (37550)
The inspectors reviewed five plant changes against the licensee's procedures and verified conformance with applicable installation and testing requirements. Accessible portions of the modifications were walked down and material condition of the surrounding areas was observed. The inspectors discussed the changes with the cognizant engineer when necessary to determine the rationale and extent of the chang Observations and Findinas Overall the modification packages reviewed were complete, of good quality and were adequate to accomplish the design changes. The Unit 1 125Vdc battery replacement modification package evidenced an appro.oriate level of pre-installation planning, given the complexity and impact of the modificabon. This included structural, testing and security requirements, such as battery room floor loading considerations, construction and operability testing consideration, and security action requirements during installation activities. Screenings and evaluations to meet 10 CFR 50.59 were included and were adequate to ensure that NRC approval was not needed prior to installing the modification. In the case of the Unit 2 battery installation, the inspectors noted that the licensee properly determined that prior NRC approval was necessary, since the Unit 2 battery replacement involved an on-line replacement. The required license amendment was submitted and approve The inspectors verified that the packages contained appropriate field instructions to ensure satisfactory installation. In the case of the essential service water linestop addition, the package identified other plant equipment that was affected by the modification and ensured good control over that modification prior to and during the linestop activities. As appropriate, the inspectors reviewed associated calculations. The inspectors determined that the calculation associated with auxiliary feedwater standpipe modification demonstrated that there was sufficient net positive suction head avai!able for the auxiliary feedwater pumps under all operating modes including at the time of supply switchover from the condensate storage tank to the essential service water system. The inspectors noted that, in the case of the emergency diesel generator Woodward governor replacement, the licensee had followed up on issues that arose I when the modification was installed at another nuclear plant and had taken steps to j
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ensure that those same problems did not occur at Braidwoo For all the modifications, the inspectors noted that post-modification testing was ,
complete and appropriate to test the extent of the modification. The testing ensured not ,
only functionality of the modification, but also ensured system operability. The post !
modification testing for the Woodward governor replacement was viewed very positively
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i by the inspectors because it identified a number of problems that occurred during the )
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I installation of the modification. For example, the wrong governor was installed because the vendor gave the licensee an incorrect part number; two failed digital reference units were installed because the parts dedication process did not identify that the vendor supplied defective digital reference units; and an incorrect actuator was installed because the maintenance staff obtained the wrong parts from stores. The licensee wrote three problem identification forms as a result of these problems and spent considerable engineering expertise in resolving them. Based on a detailed review of the identified problems, the inspectors determined that they did not result from the modification proces As a result of the licensee-identified problems with the Woodward governor modification, the inspectors followed up on the site's parts dedication process. The site staff was i struggling to assume the duties for the parts dedication process formerly performed by i the corporate staff. As part of its corrective actions, the site had included a functional test for digital reference units during receipt inspection. In addition, the licensee had investigated a trend of bad parts getting installed at the site. Corrective actions were in progress to correct the parts dedication process problem The inspectors interviewed selected site design personnel about the modifications and l
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about the reason for each modification and its effectiveness. For example, the engineer responsible for the linestop addition provided insight into the logistical challenges encountered during the installation process of that modification. The battery replacement project manager, who was also the design engineer, was extremely knowledgeable on the installation and design issues of the projec The inspectors performed walkdowns of the completed modifications to verify the adequacy of installation. A walkdown and review of the Unit 2 battery replacement '
project, ongoing at the time d the inspection, confirmed that this project was under firm control by the licensee. For example, the storage area where the temporary batteries were located was appropriately separated from other equipment on the turbine floo The inspectors also noted that there was adequate security and foreign material exclusion control over access into this area.
. . Conclusions Based on the inspection results, the inspectors concluded that the methods used to control design changes and modifications at the Braidwood plant were effective. The modification packages were adequately designed and of good technical qualit Appropriate controls were established over installation activities. Post modification testing was very good, especially in the case of the Woodward govemor replacemen E1.2 10 CFR 50.59 Evaluations and Screeninas Insoection Scooe (37001)
The methods and procedure used to control 10 CFR 50.59 safety evaluations and l
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screenings were reviewed to verify adequacy, control, and compliance with regulatory
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requirements. Emphasis in this review was on design changes and modification CFR 50.59 evaluations and screenings were discussed with cognizant licensee personnel and selected evaluations were reviewed in detail to verify implementation and compliance with the requirements of 10 CFR 50.5 Observations and Findinas implementing procedures appropriately described effective methods for controlling and performing 10 CFR 50.59 screenings and evaluat;ons. The selected 10 CFR 50.59 screenings and evaluations were verified to be appropriately prepared in accordance with the implementing procedures, and of good quality. The inspector verified through review that the screenings were appropriately performed and that no further safety evaluation was required. The evaluations adequately addressed the effects of the proposed changes on plant operations, interactions with other systems and components, any new failure modes, the effects on accidents and transients, and whether prior NRC review was required. Evaluations performed for plant changes appropriately answered the 10 CFR 50.59 questions with thorough in-depth discussions, referencing other documents as appropriate. The inspector verified that the 10 CFR 50.59 reviews were included in those submitted to the NRC under 10 CFR 50.71. Overall, the 10 CFR 50.59 evaluation and screening program was found to be effectiv Conclusions The inspectors concluded that the 10 CFR 50.59 evaluations and screenings were effective, thorough and appropriate to the plant change E1.3 Temoorary Modifications Insoection Scooe (37550)
The methods used to control temporary modifications were reviewed to verify adequacy, control, and compliance with regulatory requirements. The review included the controlling procedure and selected open temporary modification packages, which also included the appropriate 10 CFR 50.59 evaluations or screenings. Temporary modifications were discussed with cognizant licensee personne Observations and Findinas The inspectors reviewed the licensee's process to ensure control over temporary modifications. The inspectors noted that the cognizant engineer reviewed the temporary modification and completed a 10 CFR 50.59 screening or evaluation prior to installatio An action or work request was initiated for the removal of the temporary modification prior to engineering approval and issuance of the temporary modification design change package. The licensee tracked each temporary modification and performed a monthly '
review to determine if any temporary modifications were no longer needed. The inspectors confirmed that the temporary modification coordinator was completing the monthly review through review of the temporary modification monthly administrative review sheet for April 1999. Temporary modifications that exceeded their scheduled removal dates were identified for system engineering review and evaluation for
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continued need. The responsible system eng'ineer had to provide actions taken for removal orjustify extension and the expected completion date. Plant Manager approval was required to extend the scheduled removal dat The action or work request clearly indicated the temporary modification removal date so that the work could be prioritized and scheduled. The system engineer verified that the action or work request was scheduled and that all appropriate procedure change upchtes were initiated prior to signing concurrence for the temporary modification removal. The inspectors reviewed the action request for removing the temporary te:nperature monitoring instruments from the Unit 2 "B" reactor coolant pump and also verified that proper controls were taken during the removal of temporary modification 91-2-03 There were 29 open temporary modifications and only 2 had been open for longer than eighteen months. Both of the oldest open temporary modifications were scheduled for removal in May 1999. There were 17 temporary modifications on Unit 2; 14 of which were scheduled for closure during the next Unit 2 outag Conclusions Based on the inspection results, the inspectors concluded that temporary modifications were well controlled, including having stringent controls to ensure that temporary modification were removed when no longer neede E1.4 Calculations Insoection Scooe (37550)
'I he inspectors reviewed design calculations, as noted in the " List of Documents Reviewed," with an emphasis on safety and risk significant issues. Consistency among calculations, associated technical specifications, the updated final safety analysis report, and other design basis documentation was also evaluate Observations and Findinas The inspectors assessed the calculational process and content, including the identification and use of inputs, assumptions, and calculational methods. No key technical inaccuracies were identified which significantly affected the results or conclusions of a calculation. However, throughout the calculations, numerous minor inaccuracies were identified. These indicated that the attention to detail during the development, review, and approval of a calculation or revision to a calculation might not always have been sufficien The inspectors noted that many of the calculations were performed by outside organizations and ascertained that the licensee's role in reviewing and approving these calculations had not been formally controlled prior to May 1998, when Revision 6 of procedure NEP 12-02," Preparation, Review, and Approval of Calculations," was approved. The inspectors considered the procedural requirements in NEP 12-02, Revision 6, adequate to generally ensure the appropriate review and approval of
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calculations performed by outside organizations. Nevertheless, even after the implementation of NEP 12-02, Revision 6, the inspectors identified two calculations containing an unverified assumption, regarding a residual heat removal orifice plate diameter, and several other minor inaccuracies. In addition, the inspectors identified an example where an erroneous calculation methodology from an earlier calculation was used in two design calculations prepared after Revision 6 was implemented. The licensee issued three problem identification forms on these errors (A1999-00988,
-00989, and -01117).
During review of the updated final safety analysis report (UFSAR) section 6.5.2.2, against the latest emergency core cooling system net positive suction head calculations, the inspectors noted that the net positive suction head - available for the residual heat removal pumps was not specifically addressed (PlF A1999-01131). The design basis values in the calculations were otherwise found to be consistent with the values in the technical specifications and the UFSA Conclusions The inspectors concluded that, while the final numerical results were acceptable, calculations at times exhibited a lack of attention to detail based on the unverified assumptions and incorrect methodologies observed. There was good correlation among the calculations, the technical specifications, and the UFSA E1.5 Surveillances Insoection Scooe (37550)
The inspectors evaluated selected technical specification surveillances associated with pump inservice testing to verify that the design basis values were correctly translated into the testing progra Observations and Findinas l The inspectors verified that the surveillance test acceptance criteria for the residual heat removal and safety injection pumps were established consistently with both the American Society of Mechanical Engineers (ASME) Code and the updated final safety
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analysis report. The acceptance criteria were based upon the most restrictive parameter '
and were properly translated into the procedures. The inspectors verified that the recorded pump test data was appropriately evaluated against the test acceptance criteria.
1 CAn.plusions The inspectors concluded the residual heat removal and safety injection pump inservice testing surveillances ensured that both the ASME Code requirements and the design basis requirements were satisfied.
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E2 Engineering Support of Facilities and Equipment E Modification Suooort and Closeout l Inspection Scope (37550)
The methods used by engineering to provide support to the modification process were reviewed to verify adequacy and control. Additionally, the process used by the licensee for final review and closeout of modification packages for installed plant changes was tilso reviewed to verify adequacy, control and timeliness. The review included relevant procedures and records as well as discussions of the processes with cognizant licensee personnel, Observations and Findinas Licensee procedures adequately described the duties and responsibilities of the system engineers, design engineers and the engineering support staff in supporting the I modification process, including the final review and closeout of the modification packages. The licensee's modification process structure included specific interface and l communications between the key engineering disciplines (Design Engineering and i' System Engineering). The process also specified interfaces with other station departments such as Operations, Maintenance and Plant Support. For example, interviews with cognizant system and design engineers revealed that, per procedures, they were required to monitor the status of the design change package assigned to them from initiation to completion and testing. The inspectors also noted that routine planning l and status meetings were attended by individuals involved in the specific design chang ,
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The inspectors selected five modification packages for review to ensure that the design !
approval and closeout were in accordance with station procedures. Based on review of the selected modifications and interviews with the Modification Coordinator, it was ,
determined that the licensee's approval and closeout process was effective. By j procedure, design change documentation was to be closed out within 90 days from the ]
time the system was declared operable. The inspectors noted that configuration i
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. information such as control room system procedures and critical drawings were ;
maintained adequately and updated in a timely fashion for these modification package j For example, the licensee's closure trending data indicated that for March 1999, nine !
modifications were closed with seven being closed in under a week. The backlog of drawings following the Unit 1 steam generator replacement outage had decreased from in December 1998 to 102 in April 1999. The inspectors did note a few examples where the licensee had not completely filled out the administrative final closeout checklist forms . However, the inspectors independently determined that the actual activity ;
(procedure or drawing revisions, etc) listed on the checklist was appropriately accomplished. The inspectors were informed that revisions to the modification procedures implemented just prior to the inspection were expected to resolve the minor weakness in the modification package final closure proces '
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Based on the results of reviews of documentation and interviews with heensee ;
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L The methods used to perform operability determinations were reviewed to veriiy the 1 adequacy, control, and compliance with regulatory requirements. The review induded )
nine selected operability determinations that were performed in 1998 and 199 j l Observations and Findinos The implementing procedure described methods for controlling the adequacy of operability determinations. The operability determinations had sufficient information to l evaluate the issue and confirm the initial operability call made by the Shift Manager l during the associated problem identification form review. The operability determinations j were found to have been issued in a timely manner, generally within one to five days for i the initial evaluation. Many operability determinations were then followed with a more j extensive evaluation, which could take several months to complete. Review of these i extensive evaluations did not identify any problems. The inspectors noted that one deteimination of residual heat removal system operability required generation of an ;
instrument error calculation, which was included in the operability determination package. The number of operability determinations in place and amount of time that the I operability determinations were in effect did not appear excessive.
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The inspectors identified an apparent documentation weakness where an operability 3
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determination for two failed reactor coolant pump temperature detectors did not appear .
to address updated safety analysis report (UFSAR) commitments regarding operator action timeliness. Upon questioning by the inspectors, the licensee identified other parameters which would indicate incipient reactor coolant pump bearing failure in j sufficient time to satisfy the UFSAR commitments. This clarified the apparent l documentation discrepanc Conclusions The inspectors concluded that the licensee's operability determination process was effective. Adequate bases were provided for the operability determinations and the evaluations could be followed without difficult i l
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.e E2.3 Operatina Exoerience Proaram a, Insoection Scooe (37550,40500)
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The inspectors reviewed the methods of obtaining and using industry operating experience to prevent problems at Braidwood, including the operating experience program. The inspectors also reviewed industry information obtained from outside the operating experience program. The reviews encompassed the operating experience program controlling procedure, the 1998 annual operating experience program effectiveness review, a selected sample of operating experience program evaluations and a selected sample of events at other plants. The inspectors interviewed cognizant licensee personnel about the industry experience progra Observations and Findinos The inspectors determined that the operating experience program controlling procedure adequately defined the responsibilities for the receipt, dissemination, and evaluation of industry information such as NRC notifications, institute of Nuclear Power Operations (INPO) notifications and vendor information letters. The procedure also properly directed that changes to the plant or procedures be implemented if deemed appropriate by the evaluations. In addition, Braidwood personnel obtained and disseminated operating information from other sources; for example, Braidwood regulatory assurance personnel attended NRC, INPO and Nuclear Oversight exit meetings at other Comed stations. The inspectors verified that operating experience information was widely distributed across the site by a spectrum of methods including computer, newsletter, plan-of-the-day meetings, review committee meetings and "just-in-time" job and shift briefing The inspectors found the Operating Experience Program Coordinator to be experienced, diligent and adequately accomplishing the program. The inspectors identified strengths that assisted the operating experience program coordinator including strong support from the corporate operating experience progmm coordinator and the other Comed station operating experience program coordinators; Counterparts from other Comed stations routinely assisted Braidwood personnel by phone, computer, or site visits The inspectors determined that the selected sample of operating experience program evaluations were properly accomplished with appropriate plant and procedure changes being made. The inspectors selected a sample of events at other plants and found that the events that resulted in operating information notifications to Braidwood were properly captured by the operating experience program system. The inspectors found that even though some events had not resulted in operating experience program notifications,
. Braidwood personnel had still retrieved and properly dispositioned information about the events, e.g., Braidwood requested the Vectra report on the lessons leamed from the Cooper Station's installation of a new diesel generator govemor system before Braidwood installed the system. The inspectors did not identify any industry information that might have been better analyzed to prevent an event at Braidwood other than the cases already identified by the licensee in its thorough effectiveness revie b ..
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. Conclusions Based on the inspection results, the inspectors concluded that the methods used to obtain and disposition industry operating experience were effective. A strength noted was the support provided to the Operating Experience Program Coordinator by the corporate and other site coordinator E2.4 Emeroency Llahtina Unit Maintenance Rule Performance Criteria a. - Inspection Scope (40500)
The inspectors reviewed emergency lighting unit maintenance rule performance criteria, emergency light surveillances and the associated procedures. The inspectors also held discussions with the system engineer and maintenance rule coordinator regarding the emergency light performance criteri Observations and Findinas The inspectors noted that, in January 1999, the maintenance rule performance criteria for emergency lighting units had been changed from a 10 percent functional failure rate (24) per year to 15 functional failures per month. In order to determine the functional failures, the licensee included information from the monthly observations, quarterly conductivity tests, and 18-month 8-hour battery discharge surveillances. The inspectors were concemed that this change in the performance criteria could allow failure of the 8-hour discharge tests on all 242 emergency lighting units without the batteries being moved into the maintenance rule (a)(1) category. The rational behind this concem was the method used to perform the 18-month surveillance: 227 of the lighting units were !
subdivided into seventeen groups of approximately 14 units each. (The remaining 15 J were within the drywell or other inaccessible areas and were only tested during refueling outages.) On average, one group of lights was tested every month, although some months no discharge testing was performed and during other months two or three groups of batteries were tested. As the battery tests historically caused the majority of the functional failures, the revised performance criteria effectively could mask a i 100 percent battery discharge failure rate. They could also lead to the licensee concluding that problems were fixed because several months passed without further failures, when in actuality no discharge testing was performed during that tim After the inspectors questioned the performance criteria based on the above concerns, the licensee's Maintenance Rule Committee met and changed the criteria to "Less than i or equal to 15 emergency lighting unit failures per month, with no more than 22 battery discharge test failures per quarter." The inspectors further questioned these criteria as 22 battery discharge tests per quarter allowed an approximately 52 percent failure rat The licensee acknowledged that the criteria was actually set based on the quarterly conductance tests, which tested all 227 lighting units per quarter, and stated that the performance criteria would be further revised to correct the error (PIF # A1999-01526).
10 CFR 50.65(a)(2) states that monitoring, as specified in 10 CFR 50.65(a)(1), is not ,
required where it has been demonstrated that the performance or condition of a structure, system or component is being effectively controlled through the performance
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of appropriate preventive maintenance, such that the structure, system or component remains capable of performing its intended function. Contrary to 10 CFR 50.65(a)(2), the licensee had not demonstrated that the performance or condition of emergency lighting units was being effectively controlled through appropriate preventive maintenance. This was evidenced by the failure to establish performance criteria with adequate technical basis such that the emergency lighting units should have been monitored in accordance with section (a)(1) and was a violation. However, this Severity Level IV violation is being treated as a non-cited violation (NCV), consistent with Appendix C of the NRC Enforcement Policy (50-456/99005-01(DRS); 50-457/99005-01(DRS)). Conclusions The emergency lighting units maintenance rule performance criteria of 15 functional failures per month was not adequate and could have resulted in masking performance problems with emergency lighting units. A non-cited violation was issued as immediate
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corrective actions were taken and the issue was entered into the corrective action j syste l E2.5 Emeroency Liahtina Unit Surveillances Inspection Scoce (37550)
The emergency lighting surveillances were reviewed for completeness, accuracy, procedure compliance, and emergency lighting unit availabilit ;
) Observations and Findinas The inspectors identified a problem with Braidwood Operating Surveillance (BwOS) l XEL-2, " Emergency Lighting Surveillance," Revision OE3. The inspectors noted that the procedure required the maintenance staff to assess the adequacy of emergency lighting units based on the number of hydrometer discs that were floating at the top of the emergency lighting battery. The procedure considered two discs being up for a two lamp emergency lighting unit and three discs up for a 3-lamp emergency lighting unit as acceptable. Actually, the number of hydrometer disks floating actually indicated the battery charge: Three discs up meant the battery was 100 percent charged and two discs up meant that the battery was 75 percent charged. Since the batteries had a constant trickle charge applied to them, a battery with only two disks floating indicated a condition where further investigation into the battery status was necessary. The inspectors also noted several cases in completed surveillance where 3-lamp batteries had only two discs up and no corrective actions were taken. The licensee agreed that the procedure was inadequate and PIF # A1999-01249 was issued to enter these problems into the licensee's corrective action program. The inspectors also identified a number of other problems with completed surveillances such as surveillances that contained missing data,8-hour discharge tests that stopped the test at 10 to 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> with battery failure, conductance tests following an 8-hour discharge test, and batteries that were not protected during electrical bus outages. These conditions indicated that the surveillances had not been closely reviewed following completion. After the inspectors presented these problems to the licensee, PIF # A1999-01249 was expanded to review the overall procedure and take appropriate corrective action .
10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings,"
requires, in part, that instruction, procedures, or drawings include appropriate acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, as of April 23,1999, Procedure XEL-2 was inadequate because the acceptance criteria allowed potential battery problems to remain uncorrected. However, this Severity Level IV violation is being treated as an NCV, consistent with Appendix C of the NRC Enforcement Policy (50-456/99005-02(DRS);
50-457/99005-02(DRS)).
c. Conclusions The licensee failed to identify potentially adverse battery conditions due to an inadequate surveillance procedure and inadequate surveillance procedure reviews. A non-cited violation was issued as immediate corrective actions were taken and the issue was entered into the corrective action syste E4 Staff Knowledge and Performance a. Inspection Scope (37500,40500)
Throughout the inspection, routine interactions with licensee personnel occurred. During these interactions, the inspectors assessed the thoroughness of licensee staff knowledge and the effectiveness of staff performanc b. Observations and Findinas
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System engineers were found to be aware of problems on their systems and taking adequate steps to correct those problems. System engineers performed thorough root cause analysis and investigations as needed to determine the actual cause of equipment problems. Plant engineers demonstrated keen technical understanding in their areas of responsibility and understanding of plant design, procedures, organizational interfaces, and responsibilities. Specifically, the inspectors noted that the mechanical design engineers were knowledgeable and experienced in a broad range of mechanical engineering design features and specific calculations, heat exchanger performance l monitoring, and inservice testing activities. The Operating Experience Program Coordinator was found to have considerable operating experience and to be diligent in applying that experience to his position. The Event Screening Committee Coordinator and the Problem Identification Trend Coordinator significantly contributed to the success of the problem identification process, due to their high level of expertise in their assigned areas. The inspectors noted that personnel were well prepared for Plant Operations Review Committee meetings, both in regard to giving presentations and having read the technical background. Aggressive, technically probing questions were asked during these meetings. The Nuclear Oversight staff were found to be knowledgeable, qualified and dedicated. Throughout the plant, personnel were aware of the governing procedures and appeared committed to following the I i
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- Conclusions The inspectors concluded that the licensee staff was highly knowledgeable and competent regarding the issues being addresse E5 Engineering Training and Qualification
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E5.1 ' Review of Enaineerina Trainina insoection Scope (40500)
The methods used to verify and track the qualifications and training of engineers were reviewed to verify the adequacy, control, and compliance with regulatory requirement Five selected engineering personnel training records were reviewed and discussions held with cognizant licensee personnel about the training progra Observations and Findinas I
The implementing procedure described methods for controlling the adequacy of the engineering training. The training records were maintained at the onsite training cente The team reviewed five engineer training records. The training records for the engineers were found to be up-to-date. The training records that were reviewed were for experienced engineers and their training had been verified by testing or documented previous training. The training program was standardized thorough out the Nuclear '
Generation Group. The licensee had essentially completed transferring to a
- standardized system of certification guide records for the engineering training records at Braidwood. The inspectors also noted that each group had a training coordinator and that Nuclear Oversight assessed engineering training and qualifications as part of their master audit plan. Only a limited number of personnel were trained on performing 10 CFR 50.59 evaluations and root cause evaluations. Only engineers that were certified to perform 50.59 evaluations could obtain an evaluation number from the computer database. The inspectors confirmed that the individuals who performed root cause analyses had received specific root cause investigation training prior to performing
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a root cause analysi Conclusions
. The verification and the tracking of engineering training appeared satisfactory. The program, procedures and certification guides provided a sufficient level of trainin E6 Engineering Organization and Administration
- E6.1. System Enoineerina Prooram . Inspection Scope'(37550)
' The inspectors reviewed the system engineering program to ensure the system )
engineers were maintaining their systems, and providing good support to other plant organizations. The inspectors reviewed the items on the operator work-around,
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temporary modifications, open operability determinations, chronic problems, and degraded equipment lists. The inspectors interviewed system engineers, operations and maintenance workers, their supervisors, and site senior managemen Observations and Findinas System engineering provided good support to the plant by ensuring that problems were fixed, reducing the number of problem items such as those on the operator work-around and degraded equipment lists, responding to plant events and leading complex troubleshooting and repair teams, such as the responce to the April 4,1999, heater drain tank rupture disk failure. The overall system engineering backlog was being reduced, with priority established by the safety significance of the items. The system engineers were knowledgeable of their systems, including modifications and system issues. The system engineers maintained good communications with other plant organization Everyone interviewed stated that system engineers appropriately responded to an entire spectrum of requests from events requiring plant shutdown to information requests from operations personnel on backshift, with timeliness appropriately commensurate with the significance of the request. System engineers kept track of their systems, such as through the quanerly physical check for temporary modifications, both authorized and unauthorize The inspectors also evaluated system engineering support to the plant during the modification walkdowns by evaluating the material condition of the engineers' assigned systems. The inspectors did not observe any excessive leakage of water or oil; for example, the service water pump room was dry and clean. There was no indication of damage from hydraulic transients on components, pipes, or supports. The inspectors noted that equipment was clean and secured and observed that the material condition of the plant was goo Conclusions The system engineers were knowledgeable, diligently managed their systems and provided good support to the plan E6.2 Review Committee Activities Inspection Scooe (40500)
The inspectors reviewed the implementing procedures and records for the onsite and offsite review committees that review station performance and operations related to safety-related activities. The review included evaluating committee minutes, audits, and followup actions of items identified by the safety committees. The functions, findings and activities of these committees were discussed with cognizant licensee personnel, Observations an'd Findinas The inspectors noted that the key committees involved in monitoring / evaluating plant performance and safety were the Nuclear Safety Review Board (NSRB), Plant Operations Review Committee, independent Safety Engineering Group, and the Quality
Review Team. The NSRB, which met approximately every six months, functioned as a collegial body that used the professional experience and expertise of its members to advise the company senior officials on matters related to nuclear safety. Members of the NSRB for each Comed plant typically included senior corporate managers, senior site management and at least three outside senior advisors. The NSRB charter was similar for all the Comed plants. The NSRB was not a line function and did not substitute for the station line management review. It provided an independent review and oversight function separate from the daily operation of the plant. NSRB used subcommittees to screen issues to determine which items should be actually reviewed by the NSRB. For example, the independent Safety Engineering Group engineers were assigned the responsibility to review all 10 CFR 50.59 evaluations on behalf of the NSRB. Only a selected few were forwarded for review by the full NSRB, such as those which required NRC approval prior to implementation.
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Through review of the NSRBs charter, procedures, board meeting minutes and interviews with cognizant licensee personnel, the inspectors determined that the NSRB was actively and effectively performing their assigned responsibilities. For example, the inspectors noted that in December 1998, the Operation and Maintenance Subcommittee of the NSRB expressed their concerns with the reactor coolant system pressure control of the pressurizer. Since startup of Unit 1 following the last outage, the station experienced pressure control anomalies during feedwater manipulations at low power operation. The inspectors noted that the NSRB aggressively pursued this finding and recommended that the Byron and Braidwood stations address this issue. This NSRB action prompted the Braidwood station to form a team of operators and engineering personnel to perform a root cause analysis of this issue. The inspectors verified that this was being appropriately tracked in the licensee's corrective action program (NTS 456-357-98-0301).
The inspectors also assessed the activities of the Plant Operations Review Committee (PORC) and attended two PORC meetings during this inspection. The PORC was a multi-disciplined committee responsible for providing an oversight review of documents required for safe operation of the plant. All modification packages, including 50,59 evaluations, which affected nuclear safety were sent to PORC for review. The inspectors determined that the modification packages referenced in section E2.1 of this report vare reviewed by the PORC. The inspectors noted that PORC responsibilities were adequateiicutlined in detailin plant procedures. The meetings observed by the inspectors were conducted in a professional manner and the participants were well prepared, based on their aggressive participation in the technical discussion Regarding the Independent Safety Engineering Group (ISEG) and Quality Review Team activities, the inspectors noted that these groups consisted of dedicated and experienced engineers and management personnel. The Quality Review Team's function was to provide senior level management oversight of the overall quality of plant engineering products. Previously, these activities were performed by the Engineering Assurance Group. The inspectors noted that the Quality Review Team group appeared more detailed in its reviews of engineering products, based on a larger number of engineering products reviewed being retumed to the responsible engineers for revisions. The ISEG, part of the Nuclear Oversight Organization, was responsible for examining plant operating characteristics, NRC issuances, industry advisories, and other information that
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might indicate areas for improving plant safety. The ISEG performed independent reviews and audits of plant activities including maintenance, modifications, operations and programmatic requirements of plant activities. The inspectors also noted that the licensee was adequately addressing a self-identified issue (PIF A1999-00217) regarding a backlog of 10 CFR 50.59 evaluations. The inspectors were informed that during 1998, due to a weakness in the transmittal process of 50.59 evaluations, a large number of evaluations were not sent to the nuclear oversight organization for review. This resulted in a large backlog of 50.59 evaluations to be reviewed by ISEG. The inspectors confirmed that adequate actions were taken to prevent recurrence of this problem, including specifying that the Regulatory Assurance Department was responsible for ensuring that ISEG received all 50.59 evaluations. There was a backlog of approximately 125 50.59 evaluations (about three month's worth) as of March 1999. The inspectors also noted that this backlog did not impact plant operation, as it was an independent overview of the 50.59 process and not part of the line approva Conclusions The inspectors concluded that the onsite and offsite review committees were effective in oerforming their assigned reviews, investigations and evaluations. Members of these committees were aggressive in pursuing plant problems and issue E7 Quality Assurance in Engineering Activities E7.1 Corrective Action Proaram Insoection Scope (40500)
The team assessed the corrective action program through review of problem identification forms (PIFs) and interview of cognizant personnel concerning the corrective action and PlF processes. The team reviewed approximately 20 PlFs in detail to determine the licensee's effectiveness in identifying, resolving, and preventing problem Observations and Findinas A detailed review of approximately 20 PIFs determined that there was a good description of the problem, the PlFs were correctly prioritized, and corrective actions were appropriate for the significance of the problems. There was excellent engineering involvement for assessing technical issues in PIFs. The most significant PIFs had root cause determinations and effectivsness reviews performed. Overall, there was a low threshold for the generation of PIFs. Also, the licensee stated that the maintenance departments were more willing to report problems than in the past; however they acknowledged that a root cause process focus on individual performance still resulted in a much lower PlF generation rate in the maintenance area compared to other departments. The inspectors also noted some cases where the engineering personnel did not write PIFs for identified station problems. For example, the team identified that PiFs were not issued when the licensee acknowledged that the maintenance rule performance criteria for emergency lights was inaccurate, nor when reactor coolant pump temporary instrumentation faile r
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l l For each PlF being reviewed, the team requested additional information to determine
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if the licensee should have taken additional corrective actions beyond what was listed in the PlF package. Although the scope of corrective actions for some of the PlFs seemed limited, the inspectors determined that the investigations occurred and corrective actions were taken as a result of previous PlFs or other problems. For example, PIF A1998-02638 and PlF A1998-02657 were written for " loose parts alarms."
The documentation from these two PIFs indicated that relatively limited investigations had occurred and few corrective actions were taken for these problems. However, the team determined that the investigation and analysis of data were actually very thorough using an onsite expert, who gathered additional noise data for analysis, and an offsite contractor who analyzed the data. In addition, the team did not find any foreign material exclusion problems or increases in primary activity levels from core damage that would lead to the conclusion that there was a loose part in the primary system of the reacto The team noted that the senior station managers were the primary members of the Event Screening Committee. The major departments were represented and required to be present to meet quorum requirements. Approximately 25 to 50 PlFs were reviewed during daily event screening committee meetings. If necessary, the licensee staff was contacted for additional information on the issue in the PIF. The team noted that event screening committee approval was also required to extend a commitment date. This l minimized the number of extension requests. Wct PlFs problems were investigated and had corrective actions assigned in the time requi.U u procedure. If a PlF was not properly screened, it was sent back to the event screening committee to be rescreene Contributing to the effectiveness of the corrective action program was the Correction Action Review Board, which held the PlF investigations to a high standard and rejected those which did not meet its expectations, Conclusions The team concluded that the actions taken once a problem was identified were indicative of a strong corrective action program. The majority of plant problems appeared to be identified and assessed, and had corrective actions assigne E7.2 Root Cause Analysis and Trendina Inspection Scope (40500)
The team assessed the program for trending plant problems and reviewed selected trend reports. In addition, the team reviewed the root cause program and selected root cause reports. Effectiveness reviews were also evaluate Observations and Findinas The PIF trend coordinator tracked PIFs by the root or apparent causes with a complete history of PIFs maintained in a data bank. The data could be searched and sorted by codes, subjects, categories, and key words to identify trends. Upon identification of an adverse trend, the PlF coordinator would issue an adverse trend PIF. This PIF would be f sent to the event screening committee for evaluation and to determine if corrective l actions should be implemente l
1 The inspectors reviewed several trend reports and considered the reports to be of good quality with relevant information provided in the report. The trend coordinator was knowledgeable of most plant areas and diligent in identifying adverse trend A significant number of root cause determinations were performed during the past yea The licensee appeared to have devoted substantial resources in attempting to identify the basic causes for significant problems. Corrective actions were identified for each problem identified. Adequate resources were devoted to corrective actions once the causes were determined. The team did not identify any additional corrective actions that should have been taken. The root cause evaluations and associated corrective actions had helped prevent a recurrence of problems. For example, the root cause determination helped to correct a long term problem with tagging (See Section E8.1 of this report). Contributing to the quality of the root cause evaluations were the personnel ,
assigne The team reviewed a sample of effectiveness reviews. The licensee had identified corrective actions as ineffective in several instances. Additional corrective actions were assigned to these problems. The team considered the use of the effectiveness reviews as a good tool to identify when additional corrective actions were needed to prevent problem recurrence. The team also noted that 70 to 160 effectiveness reviews had been performed each year for the past few years, which contributed substantially to the licensee's strong performance in preventing recurrence of problem Conclusions i
The team concluded that root cause investigations were accurate and thorough. The licensee's trending program and effectiveness reviews contributed to identifying repetitive problems which required additional corrective actions.
E7.3 Audits and Surveillances Insoection Scope (40500)
The methods used to perform and control the Nuclear Oversight assessments were reviewed to verify adequacy, control, and compliance with regulatory requirements. The review included the controlling procedures; master audit plan; and selected 1997,1998 and 1999 assessments or audits as well as discussion with cognizant licensee personnel. Four corrective action audits and assessments were reviewed. Selected j 1998 self-assessments were also reviewe Observations and Findinas in December 1998, the quality assurance function at Braidwood was reorganized under the auspices of the Nuclear Oversight organization. As part of this reorganization, the method used to perform the required 10 CFR Part 50 Appendix B audits was changed to incorporate the requirements into a continuous assessment process. Therefore, at the time of the inspection, the licensee was using the terms " audit" and " assessment" interchangeabl .
The basis for this continuous assessment process was contained in a master audit pla The master audit plan had 4 major areas which were further divided into 13 matrix unit Each matrix contained a list of critical attributes to be assessed during each assessment ;
cycle. The critical attributes fulfilled all the requirements of the Quality Assurance l Topical Report. The inspectors learned that the licensee's program required verification I of all the critical attributes within a two year cycle, although actual verification of an j attribute could fall anywhere within a cycle. The inspectors considered the specification )
of critical attributes in the master audit plan to be a good enhancement of the quality i assurance program. The assessments were performed on a quarterly basis and had vigorous controls to ensure that critical attributes were not overlooked. The inspectors did note, however, that the continuous assessment process appeared much more limited in size and personnel qualification compared to the previous program. The licensee acknowledged that some assessments could have only one person on them, although they tried to have at least two. The Nuclear Oversight Manager also noted that their 1 commitment was only to have the assessment team leader to be a certified lead audito l The audits performed over the last two years were found to be generally thorough and conscientious assessments of the areas, with the more recent assessments appearing more thorough. The inspector reviewed assessment NOA 20-99-002, and its associated assessment plan, and found that the plan and assessment conformed to the procedures j and the master audit plan. The inspectors noted that the assessment reports generally I were issued within two weeks to a month following the assessment. The findings, which required a response, and observations from the assessment were entered into the corrective action system. However, the continued effectiveness of the program could not be evaluated because it had been in process for less than four months at the time of the inspectio The inspectors reviewed two corrective action audits performed in 1997, at which time j the audits were performed on a semi-annual frequency. The audits were found to be l satisfactory in the scope and depth of the evaluation. Previous audit findings were followed up to verify the corrective actions taken by the plant were still effective. The I inspectors considered the reviewing of closed findings during an audit as an effective !
method to assess the adequacy of the corrective actions and noted it was carried over into the assessment process. The 1999 program provided for an assessment of corrective actions as a critical attribute for each of the 13 matrix areas, which was to be performed only once each two year Nuclear Oversight performed an operations corrective actions assessment in January 1999. The inspectors found the assessment to be an in depth evaluation of the attribute with a number of good observations provided in PlF A1999-00368. The inspectors followed up on the corrective action taken for an item identified in the assessment and found the corrective actions taken to be appropriat The inspectors reviewed two department performance reviews and two focus area self-assessments. The inspectors determined the licensee's assessments of system engineering and operating experience program effectiveness were a strength because they clearly and comprehensively identified problems, showed areas for improvement and recommended corrective actions. Problems identified by the assessments were entered into the corrective action program by initiating problem identification forms. The
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self-assessments program was generally highly thought of by plant management as a helpful tool to identify and solve problem Conclusions The inspectors concluded that the assessment process used by the licensee's independent oversight group met the requirements of 10 CFR Part 50, Appendix B. The continuous assessment process provided an acceptable review of the area being assessed. The self-assessments program provided sufficiently in-depth information so that the assessed group could make the appropriate corrections in the areas of weaknes E8 Miscellaneous Engineering issues E (Closed) Violation 50456-96006-01: 50457-96006-01: Failure to Follow Procedures:
Deficiency Tags Not Removed for Completed or Canceled Work.' Only part "d" of this violation remained open. This violation identified several examples where work had been either completed or canceled, but the deficiency tags associated with the action requests had not been removed. The tagging problem continued to be a problem even after corrective actions for this violation were implemented. After several audits and effectiveness reviews flagged the continuing problem, the licensee took additional corrective actions, which appeared effective as a review of problem identification forms for the past year indicated that only two tagging problems occurred during this time period. The corrective actions appeared appropriate to prevent most tagging problem This violation was closed based on the reduction in the tagging problems, the planned corrective actions and the fact that the issue is captured in the corrective action progra This violation is close E8.2 (Closed) Unresolved item 50456-96016-01: 50457-96016-01: Licensee Interpretation of Design Basis for Motor Operated Valve Hot Shorts. The inspectors had questioned the licensee's position that, except for high-low pressure interfaces, the Byron and Braidwood safety evaluation report commitments for safe shutdown were approved considering only one spurious actuation, without need to account for mechanical damage to multiple valves. Because this position did not appear to be in accordance with current NRC policy but would require a change to the plant's licensing basis, the inspectors forwarded the issue to the Office of Nuclear Reactor Regulation (NRR) for review. If necessary, when a response is received from NRR, this issue will be revisited. This item is close E8.3 (Closed) Violation 50456-97012-2a: 50-457-97012-2a: Inadequate 10CFR 50.59 Evaluation. This violation was written as a result of a failure by the licensee to reference
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a discussion of ASME Code requirements for relief valves in the 10 CFR 50.59 evaluation for a temporary alteration which gagged a safety injection relief valve closed
. when it lifted prematurely during a surveillance test. The safety injection relief valve was replaced and satisfactorily tested during the A2R06 refueling outage. The inspector reviewed the response to the violation and verified that adequate actions were take This item is close a
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E8.4 (Closed) Inspection Followuo item 50-456-98022-02: 50-457-98022-02: Review of Fire Protection When Welding, Cutting, Grinding, or Performing Open Flame Work (Hot Work) Procedure. Braidwood Administrative Procedure (BwAP) 1100-15. " Fire Protection When Welding, Cutting, Grinding, or Performing Open Flame Work (Hot Work," Revision 10, did not implement some portions of the National Fire Protection Association (NFPA) Standard 51B-1984. The licensee changed the procedure to reflect the NFPA requirements. This item is close E8.5 (Closed) LER 50-457/97005-00/-01/-02: Violation of Technical Specifications Due to a Safety injection (SI) Relief Valve Lifting and Failing to Reseat. During operational testing, safety injection relief valves lifted and failed to reseat due to a limited margin between the safety injection pump start pressure transient and the valve setpoints. The licensee determined that the safety consequences of the event were minima Corrective actions included procedure revision, setpoint tolerance evaluation, testing methodology review, and replacement of the failed relief valves. The licensee has completed all committed actions, with the exception of completion of an effectiveness review. This final action is being tracked by the licensee's corrective action program (AR No. 00003665-05). The inspector reviewed the response to this minor error and verified that adequate corrective actions were taken. Although this was a violation of the technical specifications it is not being cited, consistent with Appendix C of the NRC Enforcement Policy (50-457/99005-03(DRS)). This item is close l E8.6 (Closed) LER 50-456/97009-00/-01: Potential Failure of Westinghouse Fuel Rod Design !
Criteria. The LER documented a Westinghouse identified calculational erro Westinghouse considered the problem to be generic to all plants using Westinghouse fuel, and reported it to the NRC in late October,1997. Westinghouse performed a plant specific analysis for Braidwood which confirmed that fuel failure would not occur, based on normal fuel cycles. The Westinghouse analysis was reviewed by the licensee, and the maximum fuel burnups were determined to be covered by the Westinghouse !
analysis. The licensee reported the results to NRC in the supplemental LER. The licensee has completed all committed actions, with the exception of completion of an effectiveness review. This final action is tracked by the licensee's corrective action program (AR No. 00003090-01). The inspectors had no further concerns, and this item is close E (Closed) LER 50-456/98001-00: Failure to Test Engineered Safety Feature Logic Circuit Due to Oversight by initial Review. The licensee discovered that a set of three contacts had not been tested to ensure the contacts could energize the Train "A" safeguards actuation relay while the Unit i bus was cross-tied to the Unit 2 bus. The inspectors verified that all required surveillance testing was completed. This item is close E8.8 (Closed) LER 50-456/98004-00: Main Steam Safety Valves (MSSVs) Tested in Excess of Required Setpoint Due to Suspected Metallic Bonding. During setpoint verification testing of Unit 1 main steam safety valves, five valves lifted in excess of their setpoints by greater than the three percent technical specification tolerance. The licensee's root cause investigation suspected metallic bonding between the disc and nozzle seats caused by differences in the coefficient of expansion between the disc and the nozzl Corrective actions included evaluation of recent valve test data, which concluded that the acceptance criteria for the applicable updated final safety analysis report accident
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scenarios were not exceeded, rebuilding of the five valves, and a revision to the rebuild I procedure. Additionally, the licensee planned to review past rebuild packages at both Byron and Braidwood to assess differences and their potential effect on the main steam safety valve failures. The licensee also planned to issue a supplement to the LER. These items are being tracked in the licensee's corrective action program (AR No. 00003094-03). The inspector reviewed the response to the event and verified that adequate corrective actions were taken. Although this was a violation of the i technical specifications it is not being cited, consistent with Appendix C of the NRC Enforcement Policy (50-456/99005-04(DRS)). This item is closed E8.9 (Closed) LER 50-456/98006-00: Failure to Test Contacts as Required by Technical Specifications Due to inadequate Surveillance Procedure Development Prior to initial {
Plant Start-Up. During the Generic Letter 96-01 reviews, the licensee identified portions (
of two circuits that were not being adequately tested. A contact associated with the I nuclear channel test card for the refueling water storage tank level instrument loop input )
to the solid state protection system was not verified to retum to its original state after performance of the quarterly analog channel operational test. The significance of this issue was minimal since the closed position of the contacts was verified once every six j months during response time testing. The inspectors determined that all required testing to verify proper contact operation was completed by the licensee. The inspectors also noted that a revision to the appropriate operating surveillance ( BWOS 7.1.2.1.b.1(2) )
was scheduled for completion on 5/1/99. Although this was a violation of the technical ,
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specifications it is not being cited, consistent with Appendix C of the NRC Enforcement Policy (50-456/99005-05(DRS)). This item is close E8.10 (Closed) LER 50-456/98007-00: Non-Conservative Error Detected in Vendor's Analysis :
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Code. The LER documented a Westinghouse calculational error that occurred due to a failure to replace a generic value with Braidwood/ Byron specific data. Following evaluation of several options, the licensee determined the optimum solution would be to replace the inaccurate boron dilution prevention system with volume control tank level monitoring and subsequent operator actions. The licensee determined that this solution ,
would require a license amendment and was proceeding with the required action, including submittal of the change to NRC. The inspectors concluded that the licensee's actions to date were acceptable, and that further actions would be controlled through the corrective action process (AR No. 00003097-02). This item is close IV. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management in an exit meeting on April 23,1999. The inspectors noted that no documents provided during the inspection were identified as proprietary. The licensee acknowledged the information presented and agreed that no proprietary information was provided to the inspector . )
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I PARTIAL LIST OF PERSONS CONTACTED Licensee M. Cassidy, Regulatory Assurance - NRC Coordinator J. Kuchenbecker, System Engineering Manager T. Luke, Engineering Manager J. Nalewajka, Assessment Manager M. Riegel, Nuclear Oversight Manager K. Schwartz, Station Manager T. Simpkin, Regulatory Assurance Manager T. Tulon, Site Vice President R. Wegner, Operations Manager NRG l
C. Phillips, Senior Resident inspector ]
IDNS l
J. Roman lNSPECTION PROCEDURES USED
IP 37001: 10 CFR 50.59 Safety Evaluation Program l I
IP 37550: Engineering IP 40500: Effectiveness of Licensee Controls in identifying, Resolving and Preventing Problems IP 92700: Onsite Followup of Written Reports of Non Routine Events at Power Reactor Facilities IP 92903: Followup - Engineering l l
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ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-456/99005-01 NCV Inadequate Maintenance Rule Performance Criteria-50-457/99005-01 50-456/99005-02 NCV Inadequate Emergency Lighting Unit Surveillance Procedure 50-457/99005-02 50-457/99005-03 NCV Technical Specification Violation - LER 457/97005 50-456/99005-04 NCV Technical Specification Violation - LER 456/98004 50-456/99005-05 NCV Technical Specification Violation - LER 456/98006 Closed 50-456-96006-01 VIO Failure to Follow Procedures 50-457-96006-01'
50-456-96016-01 URI - Licensee Interpretation of Design Basis for Motor Operated Valve 50-457-96016-01 Hot Shorts 50-456-97012-02 VIO Inadequate 10CFR 50.59 Evaluation 50-457-97012-02 50-456-98022-02 IFl inadequate Fire Protection Procedure 50-457-98022-02 50-456-99005-01 NCV inadequate Maintenance Rule Performance Criteria 50-457-99005-01 '
50-456-99005-02 NCV Inadequate Emergency Lighting Unit Surveillance Procedure
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50-457-99005-02
.50-457/99005-03- NCV Technical Specification Violation - LER 457/97005 50-456/99005-04 NCV Technical Specification Violation - LER 456/98004 50-456/99005-05- NCV Technical Specification Violation - LER 456/98006
- 50-457/97005-00 LER Violation of Technical Specifications Due to an Safety injection 50-457/97005-01- Relief Valve Lifting and Failing to Reseat 50-457/97005-02 50-456/97009-00 LER Potential Failure of Westinghouse Fuel Rod Design Criteria 50-456/97009-01 50-456/98001-00 LER Failure to Test Engineered Safety Feature Logic Circuit Due to initial Review Oversight 50-456/98004-00 LER- Main Steam Safety Valves Tested in Excess of Required Setpoint Due to Suspected Metallic Bonding l 50-456/98006-00 LER Failure to Test Contacts as Required by Technical Specifications !
50-456/98007-00 LER Non-Conservative Error Detected in Vendor's Analysis Code j
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LIST OF ACRONYMS USED ASME American Society of Mechanical Engineers BwAP Braidwood Administrative Procedure BwOS(R) Braidwood Operating Surveillance Procedure BwVS(R) Braidwood Engineering Surveillance Procedure CWPl Common Work Practice instruction CFR Code of Federal Regulations DRS Division of Reactor dsfety
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FCS Field Change Screening gpm Gallons Per Minute INPO Institute of Nuclear Power Operations ISEG Independent Safety Engineering Group ITS Improved Technical Specifications LER Licensee Event Report MDM Mechanical Department Memo NCV Non-cited Violation NEP Nuclear Engineering Procedure NFPA National Fire Protection Association NOA Nuclear Oversight Assessment NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSP Nuclear Station Procedure NSRB Nuclear Safety Review Board PlF Problem identification Form PORC Plant Operations Review Committee PTES Procedure, Test, or Experiment Screening RHR Residual Heat Removal QAA Quality Assurance Audit QAS Quality Assurance Surveillance SI Safety injection TSC Technical Support Center UFSAR Updated Final Safety Analysis Report Vdc Direct Current Voltage
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LIST OF DOCUMENTS REVIEWED The following is a list of licensee documents reviewed during the inspection, including documents prepared by others for the licensee. Inclusion on this list does not imply that NRC inspectors reviewed the documents in their entirety, but, rather that selected sections or portions of the documents were evaluated as part of the overall inspection effort. Nor does inclusion in this list imply NRC acceptance of the document, unless specifically so stated in the body of the inspection repor Assessments and Audits QAA 20-97-03 Corrective Actions I, March 14,1997 QAA 20-97-10 Corrective Actions ll, September 12,1997
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QAA 20-98-01 Quality Audit Report, Operations, November 16,1998
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Comed-98-03 Emergency Preparedness / Radiological Environmental Monitonng Program / Offsite Dose Calculation Manual Audit Report, May 15,1998 Comed-98-04 Maintenance Audit, June 24,1998 Comed-98-05 Performance Review Meeting and the Nuclear Generation Group Engineering Audit, July 27,1998 QAS 20-98-002 Controlling Switchyard Work and Impact on Nuclear Safety, Revisioni QAS 20-98-003 Implementation of High Risk Activities OAS 20-98-004 Temporary Alteration 98-2-002 for Control Rod C-11 QAS 20-98-005 Engineering Assessment Prior to AD inspection QAS 20-98-008 Operating 'C' Simulator Performance QAS 20-98-009 Unit 2 Forced Outage OAS 20-98-010 Licensed Operator Medical Qualifications OAS 20-98-011 Review of Braidwood Action Plan Regarding LaSalle Service Water Event, April 8,1998 QAS 20-98-012 Assessment of Component Cooling Heat Exchanger Evolution, May 7,1998 QAS 20-98-014 Evaluation of Timeliness of Retuming Reactor Protection System and Engineered Safety Feature Bistables to Operable Status QAS 20-98-015 Security Assessment, June 4,1998 QAS 20-98-017 Assessment of Compressed Gas Cylinder Storage, June 2,1998 OAS 20-98-018 Environmental Qualifications QAS 20-98-024 Diesel Generator Jacket Water Temperature Contro!
QAS 20-98-028 Pre-Outage Shutdown Safety QAS 20-98-030 Pre INPO Assessment QAS 20-98-039 Control Room Logs. August 26,1998 QAS 20-98-040 Strategic Reform Initiative #2 - Upgrade Operations Department Leadership Role in Ensuring Excellent Plant Operations QAS 20-98-041 Steam Generator Replacement Project Replacement Steam Generator Weld Prep Machining Reactor Coolant and Main Steam Nozzles, September 1,1998
' QAS 20-98-042 Environmental Qualifications Program, Revision 1 I
QAS 20-98-043 NGG-3: Excellence in Material Condition QAS 20-98-046 The Maintenance Rule Program l
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QAS 20-98-048 Process Controls for Nuclear Generation Group Corporate Procedures, October 30,1998 QAS 20-98-049 Unit 1 & 2 Steam Generator Moisture Carryover Test, January 11,1999 QAS 20-98-052 Y2K Readiness QAS 20-98-053 Improved Technical Specifications QAS 20-98-054 Proper Entry for Limiting Condition for Operation Action Requirements QAS 20-98-055 Closed Cooling Chemistry and Chemistry Technician Knowledge, December 18,1998 QAS 20-98-056 Radioactive Material Control, January 26,1999 QAS 20-98-057 Radiation Protection Records Control, January 26,1999 QVL 20-98-030 Nuclear Oversight Affirmation of Braidwood Unit 1 Restart Readiness Following A1R07 NOA-20-99-001 Plan for Assessment of Operations - Status Indications, December 14,1998 NOA-20-99-001 Operations - Status indication, January 20,1999 NOA-20-99-002 Operations - Corrective Actions, February 8,1999 Corrective Actions Department Quarterly Self Assessment, October 16,1998 Engineering Self Assessment in Preparation for NRC Engineering and Technical Support inspection, dated March 11,1999 Engineering Self Assessment in Preparation for NRC Architect / Engineer Design inspection, March 7,1998 Engineering Second Quarter Self Assessment of Design Change Testing, July 7,1998 Regulatory Assurance Fourth Quarter Performance Review, January 28,1999 Braidwood Pre-Outage Assessment (A2R07), March 15,1999 Calculations ATD-0111 Maximum Containment Flood Level, Revision 13 BRW-96-438-M Determination of the Maximum Containment Recirculation Sump pH Post Loss of Coolant Accident, Revision 5 BRW-98-0384-M Safety injection Pump Recirculation Flow Deadhead Calculation, i Revision 0 BRW-98-0100-M Containment Samp Zone of influence for Failed Coatings, Revision 2 BRW-98-1032-1 Residual Heat Removal (RHR) ECCS Pump Flow & Pressure Accuracy Evaluation, Revision 0 BRW-98-1081 RHR Pump ASME Surveillance Flow Accuracy Evaluation, Revision 0 CS-5 Net Positive Suction Head Available for RHR & Containment Spray Pumps, Revision 3 ,
HELB-32 Containment Sump Blockage Due to the Postulated Failure of Undocumented / Unqualified Coatings, Revision 0 NED-l-ElC-0081 Containment Floor Drain Sump Water Level Channel Error Analysis, Revision 2 NED-I-EIC-0082 Containment Floor Water Level Channel Error Analysis, Revision 3 NED-I-EIC-0141 Diesel Oil Storage Tank Indication Accuracy at Normal Operating Conditions, Revision 1
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PSA-B-97-14 Evaluation of New Condensate Storage Tank Technical Specification Levels for Byron and Braidwood Stations, Revision 0 SI-90-01 Minimum WaterVolume Available for Containment Recirculation Sump Flooding, Revision 0 SI-90-01, App, A Minimum Containment Flood Level, Revision 8 32-1266251-00 RHR Cooldown Time, Revision 0 Effectiveness Reviews
'.9701810 Evaluation of Leaving Hot-Tap Rig in Place For Header 1MS06BB-42
,9801651 Change Set point of OPS-W449 & OPS-W450 to +%"
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Modifications E20-0-96-289-001 ~ Stop Valve Locations for Component Cooling Water Heat Exchanger Isolation E20-0-97-298 Containment Sump Screen Structure Top Surface Gaps Closure J E20-1-97-210 . install New Reactor Vessel Level Instrumentation System Probe E20-1-97-268 Expand Reactor Coolant Pump #1 Seal Leakoff Flow Transmitters E20-2-97-274 Seal Electncal Penetration into 2B Main Steam Isolation Valve Motor M20-1-94-005 Emergency Diesel Generator Govemor Upgrade, DCP 9400152 M20-1-96-001 125Vdc Battery and Rack Replacement Modification M20-1-97-002-A Unit 1 Auxiliary Feedwater Header Standpipe Extension M20-1-97-002-8 Unit 1 Condensate Storage Tank Height increase Nuclear Tracking System Entries
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456-201-98-CAQS02790 Poor Human Factoring on Breaker Configuration 457-201-98-CAQS00544 Unit 2 Reactor Coolant Pumps "A" and "C" Bearing Temperature Sensors -l
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456-201-98-CAQS03034 No Work Request Written to Repair Furmanited Drain Line 456-201-98-CAQS02865 1 A Diesel Generator Roll up Door Out-of-Service Violation 456-201-98-CAQS02934 - Digital Reference Unit Found Failed During Diesel Generator
' Govemor Modification Testing 456-200-98-CAQS00014 Completed or Canceled Action Request Tags Still Hanging in Field 456-201-98-CAOS02110 Inadequate Low Differential Pressure Alarm Setpoint for the Technical Support Center (TSC)
456-201-98-CAOS02494 Failure to Maintain TSC Emergency Makeup Air Filter Flow Recorder
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Operability Determinations- 97-163; Main Feedwater Check Valves (FWO79s), December 19,1997 98-002 Temporary Local Leak Rate Test Structures in Auxiliary Building on El 426, January 9,1998 -
98-00 Safety injection Check Valves 2SI-8819 A thru D, January 26,1998 98-009 Valves 2FWO79 A, B and C, February 5,1998 !98-010 Valve 1FWO79A- D, February 2,1998 1 98-017 - Safety injection Accumulators 1/2Sl04T A - D, February 24,1998 30 l
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.98-033 Auxiliary Feedwater Valves 1 AF005 E - H, June 2,1998 98-040 2A Residual Heat Romoval Pump Differential Pressure, July 23,1998 98-043 Reactor Coolant Pump Temperature Sensor Failure, Revision 1 99-001 Pressurizer Heaters, January 11,1999 Part Evaluations I
NEP-11-01 Evaluation of Woodward Governor Parts NEP-11-01 Evaluation of Digital Reference Unit, Sl815F73 Failed Receipt inspection List, March 1998 to March 1999 MAR Log, February to April,1999 Problem Identification Forms (PlFs)
A1997-02816 Unit 2 Loose Parts Monitoring Alarm When 2B Reactor Coolant Drain Tank Pump
. Starts A1997-02825 Unit 2 Loose Parts Failure Alarm A1997-02972 Unit 1 Loose Parts Monitoring System Spurious Alarms A1997-03017 Loose Parts Monitor Alarm A1997-03835 Loose Parts Alarm A1997-03872 Unit 1 Loose Parts Alarm With No First Out, No Tape Recording A1997-03898 Loose Parts Alarm A1997-03899 Unit 2 Loose Parts Alarm A1997-04392 Loose Parts Monitoring System Spurious Alarm A1997-04970 Loose Parts Systems Spurious Alarms With Unit 2 Heatup !
A1997-05368 Unit 2 Loose Parts Alarm A1997-05374 Unit 2 Loose Parts !
A1997-05581 Unit 2 Loose Parts Monitoring Alarm A1997-05623 Unit 2 Loose Parts Alarm A1998-00748 Loose Parts Monitoring Low Alarm l I
A1998-00918 Loose Parts Alarm A1998-01275 Potential Trend C-Team / Material Management issues A1998-01450 Unit 2 Loose Parts Alarm A1998-01492 PlFs not Written on incoming NONs CAR 20-98-049 A1998-01793 Potential Adverse Trend in Configuration Control
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A1998-02322 Foreign Material Exclusion Concem at Unit 1 Electro-hydraulic Skid A1998-02384 Effectiveness Review on Procedure Adherence 456-230-96-02800 A1998-02484 Unit 2 Alarms With Lighting and Storms Moving Through Area A1998-02631 Scaffold Attached to Safety Related Structure A1996-02638 Loose Parts Alarm A1998-02641 Diesel Generator Jacket Water and Lubrication Oil Cooler Performance Data Discrepancy A1998-02647 Unexpected Condition During Performance of Surveillance SPP 98-023 A1998-02657 Unit 2 Loose Parts Alarm A1998-02658 Unit 2 Reactor Coolant Pumps "A" and "C" Bearing Temperature Sensors A1998-02698 Inadequate Low Differential Pressure Alarm Set point For The Technical Support Center (TSC)
A1998-03011 1B Diesel Generator Fuel Strainer Inlet Piping Leak
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A1998-03014 Scaffold Work Packages Lack Sufficient Erection Information A1998-03069 Failure to Maintain TSC Emergency Makeup Air Filter Flow Recorder A1998-03089 Maintenance Rule Corporate Assessment Findings - Braidwood not Writing PlFs for All Functional Failures, September 9,1998 A1998-03209 Quality Control (QC) Identified Unsecured Mobile Item During Surveillance A1998-03316 Auxiliary Feedwater 1 AF017A Freeze Seal Failure A1998-03361 incorrect Governor Installed During Diesel Generator Governor Modification A1998-03362 Operators Required to Operate Equipment Without Proper Training A1998-03375 Incorrect Diesel Generator Governor Unit Installed A1998-03386 Poor Human Factoring on Breaker Configuration A1998-03446 Check Valve 0WS2878 Failure to close - Block of Wood A1998-03458 Failure of Seal Dam on a Reactor Coolant System Cold Leg During Media Retrieval Process A1998-03570 1 A Diesel Generator Roll up Door Out-of-Service Violation A1998-03601 Torque Wrench Out of Calibration A1998-03614 Digital Reference Unit Found Failed During Diesel Generator Governor j Modification Testing l A1998-03615 Wrong Mechanical Actuator Installed During Diesel Generator Govemor
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Modification A1998-03744 No Action Request Written to Repair Furmanited Drain Line A1998-03765 Discrepancy Between Documents and As-built 4 A1998-03875 Foreign Materialin Refuel Cavity A1998-04013 QC Identified Failed Surveillance - Design Change Made Without Design Change Controls A1998-04061 Non-Environmentally Qualified Tape Used For Environmentally Qualified Termination A1998-04161 Emergency Lights Fail Conductance Test A1998-04216 Loose Parts Monitor Alarm A1998-04404 Negative Trend Analyzed in Maintenance-Related Rework Database A1998-04425 Unit 1 Loose Parts Alarm / Noise A1999-00037 Safety Analysis Modeling of Pressurizer Heaters, January 7,1999 A1999-00063 Loose Parts Sensor Trouble i A1999-00088 1FT-RF008 Indicated Flow Greater than 1 gpm While the Sump Flow Was Less than 0.03 gpm A1999-00137 BwAP 100-15 Hot Work Procedure Missing National Fire Protection Association Required Statement - Equipment Being Used in Satisfactory Operating Condition and Good Repair A1999-00217 Nuclear Oversight identified Failure to Follow Quality Assurance Manual for Transmittal of 50.59s to Nuclear Oversight, January 25,1999 A1999-00220 Potential Trend - Work Package Quality - 17 PIFs Trend investigation A1999-00431 Failed Receipt Inspection - Vendor is Woodward Governor A1999-00628 Unit 1 Loose Parts Monitor Alarm A1999-00858 Quality Review Team Review of BRW-SESV-1999-125, March 30,1999 A1999-00889 Operating Experience (OPEX) Review of NON LS-99-010[AC] Increase in Radiation Levels - Fuel Leak, April 2,1999 A1999-00890 OPEX Review of NON BY-99-016(AB]: Loop Stop Isolation Valve Failure, April 2,1999 A1999-00891 OPEX Review of NON DR 99-016[XX]: Unexpected Transfer of Water from T113 to Vault 2 Floor, April 2,1999
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A1999-00976 OPEX Review of NON BY-99-018[DF]: Refuel Machine Failure, April 9,1999 A1999-00988 Minor Errors identified by the NRC (E&TS) in Mechanical Calculations, April 5,1999 A1999-00989 Minor Errors identified by the NRC (E&TS) in l&C Calculations, April 5,1999 A1999-01005 OPEX Review of NON Zi-99-03[XX]: Exempt Radiation Sources Not Controlled '
per Procedure, April 13,1999 A1999-01011 OPEX, Review of NON BY-99-017 R1[EB]: Energized 4kV Bus Duct inadvertently Opened, April 13,1999 A1999-01049 Failure to Identify Emergency Light Surveillance Criteria Not Met A1999-01117 Errors identified by the NRC (E&TS)in Mechanical Calculations, April 22,1999 A1999-01131 Omitted information in the Updated Final Safety Analysis Report for the Residual Heat Removal Pumps Net Positive Suction Head, April 22,1999 A1999-01136 OPEX Effectiveness Review of December 1998 - Respond to Recommendations A1999-01249 Potential Maintenance Rule Violation Identified During the E&TS Inspection, April 29,1999 A1999-01526 Maintenance Rule Performance Criteria with incorrect Verbiage, May 10,1999 Procedures BwAP 1100-23 Seismic Housekeeping Requirements for the Temporary Storage of Materials in Category 1 Areas, Revision 1 BwAP 1110-15 Fire Prevention When Welding, Cutting, Grinding, or Performing Open Flame Work (Hot Work) Revisions 6,7,10, and 10E1 BwAP 1205-4 Braidwood Station 10 CFR 50.59 Processing, Revision 7 BwAP 1300-6T1 Unit 2 Main Condenser Vacuum Indication improvement, Revision 4E1 ,
BwAP 1340-14 Critical Drawing Control, Revision 7 i BwAP 1340-15 Station Document Change Control BwAP 1600 -1 Action / Work Request Processing Procedure, Revision 41E1 BwAP 1610-3 Field Change Requests, Revision 5 BwAP 1610-5 Development of Modification Tests, Revision 5 BwAP 2321-12 - Plant Modifications, Revision 8 BwAP 2321-17T1 Setpoint/ Scaling Change Request, Revision 3 BwAP 2321-21 Exempt Changes, Revision 8 BwAP 2321-25 Plant Design Changes, Revision 4 BwHS 4002-037 Safe Shutdown 8-Hour Battery Operated Emergency Light Surveillance (Discharge Test), Revision 7 BwOS XEL-3 Auxiliary Building Unit 1, Emergency Lighting Surveillance, Revision 2 BwOS XEL-4 Auxiliary Building Unit 2, Emergency Lighting Surveillance, Revision 2 BwOSR 3.5. Emergency Core Cooling System Mechanical . Position Stop 18 Month Surveillance, Revision 0 BwOSR 3.8.1.2-1 1 A Diesel Generator Operability Monthly and Semi-Annual Surveillance, Revision 0 BwVS W-1 Technical Support Center Ventilation System HEPA & Charcoal Filter Performance Test, Revision 2 j
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BwVS W-2 Technical Support Center Ventilation System Carbon Sample Removal /
Analysis, Revisions 0 and 1 BwVSR 3.5. Visual Surveillance of Containment Recirculation Sumps, Revision 0 CAP-2 Coding and Trending Instructional Guide, Revision 0 CWPI-NSP-AP-1-10 Operating Experience, Revision 0 CWPI-NSP-TQ 1-17 Conduct of Training Manual; Engineering Support Personnel Training Program Description, Revision 0 MDM 200-6 Guidelines for Dealing With Furmanite NEP-04-00 Roadmap - Design Changes, Revision 4 a NEP-04-01 Plant Modifications, Revision 6 {
NEP-04-04 Walkdowns, Revision 1 )
NEP-04-05 Design Change Acceptance Testing, Revision 0 NEP-08-02 Field Change Requests, Revision 2 NEP-11-01 Procurement and Use of Items For Repair and Replacement of Safety Related, Regulatory Related, and Non Safety Related Equipment, Revision 4 NEP-12-02 Preparation, Review and Approval of Calculations, Revision 6 NO-16 Conduct of Off-Site Review, Revision 9 l
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NO-35 Nuclear Oversight Continuous Assessment Process, Revision 1 NO-38 Nuclear Oversight Master Audit Plan, Revision 1 l
NO-39 Independent Review of 10 CFR 50.59 Safety Evaluations, Revision 0 l
NSP-AP-1002 Plant Operations Review Committee, Revision 1 i Corrective Action Program Process, Revision 1
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NSP-AP-1004 NSP-AP-2004 Corrective Action Program Process Roles and Responsibilities, Revision 1 NSP-AP-3001 Independent Technical Review, Revision 0 NSP-AP-3004 Corrective Action Program Process Manual, Revision 1 NSP-AP-3009 Self-Assessment Program, Revision 0 NSP-AP-4004 Corrective Action Program Procedure, Revision 0 i NSP-CC-3001 Operability Determination Process, Revision 0 NSP-CC-3005 10 CFR 50.59 Safety Evaluation Process, Revision 0 NSP-CC-AA-202 Quality Review Team, Revision 0 NSP-RA-3001 Conduct of the Nuclear Review Board, Revision 0 NSWP-A-06 Operating Experience, Revision 0 NSWP-A-21 Temporary Modifications, Revision 0 NSWP-A-24 Station Scaffold Erection and Inspection, Revision 0 NSWP-A-25 Station Scaffold Installation / Modification and Removal Request, Revision 0 NTAFT ENG01 Engineering Support Personnel Training Evaluation Form, Revision 00
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NTAFT JENGxx Engineering Support Personnel Training Criteria and Certification Forms, Revision 00 Root Cause Reports
- Adverse Trend in the Number of Operational Configuration Occurrences due to Deficient Use of Error Prevention Techniques and Insufficient Communication of Expectations, May 15,1998 Adverse Trend in Work Package Preparation and Review due to inadequate Verification, Interface Among Departrnents, and Management Planning and Implementation of Process Change, January 26,1999 ' )
Door SD-172 Found Propped Open Resulting in an impairment of the Main Control Room Envelope Heating Ventilation and Air-Conditioning Boundary due to Poor Assumptions, August 12,1998 A Lack of Procedure Adherence Monitoring and Active Reinforcement of Standards by Senior Management, December 7,1998
- Action Request Deficiency Tags Not Removed After Work Completed or Canceled due to Station Management Failure to Enforce All Procedure Requirements, May 12,1998 Unit 1 Heater Drain Tank Rupture Disk Failure, January 4,1999 .
' February 1999 Trend Analysis Report, undated Trend Report, Potential Adverse Trend in Work Package Quality, January 4,1999 i
Seal Dam Failure, Blast Media Entered Piping System, NCR BD-98-070 Surveillances (Completed Work Requests) i 970118588 01 Thermal Performance Test of the Unit 0 Component Cooling Heat Exchanger, September 2,1998 980088740 01 ASME Surveillance Requirement for 1 A Safety injection (SI) Pump, November 17,1998 980104812 01 ASME Surveillance Requirement for 1B Sl Pump, December 11,1998 980122320 01 ASME Surveillance Requirement for 1 A S1 Pump, February 8,1999 980130828 01 ASME Surveillance Requirement for 1B S1 Pump, March 3,1999 980134112 01 ASME Surveillance Requirements for Residual Heat Removal Pump 1RH01PA, March 8,1999
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91-2-038' Temperature Monitoring instruments 93-2-019 Installation of Temporary Cooling to Unit 2 Control Rod Drive Cabinets 98-0-015 Installation of Jumper and Resistor Across the High Motor Slot Temperature ,
Resistance Temperature Detector Terminals l l
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98-2-004 = Installation of Ring Adaptors on Bearing Cover Fasteners to Permit Furmanite injection-98-2-008- Installation of Adaptors on Valve to Permit Furmanite injection
'99-2-004 Installation of Three Diesel Powered Air Compressors 10 CFR 50.59 Evaluations BRW-SE-1997-1655 - 125Vdc Battery Replacer ^nent, April 30,1998 .
BRW-SE-1997-1920 Routing of Hydrolaze Hose Within Auxiliary and Containment Buildings, January 4,1998 BRW-SE-1998-0033 Installation of Linestop Fitting in Essential Service Water Supply l- Line, March 13,1998-BRW-SE-1998-0038 M20-1-97-002A, Unit 1 Auxiliary Feedwater Header Stand-Pipe Extension, January 16,1998 BRW-SE-1998-0039 M20-1-97-002-B, Unit 1 Condensate Storage Tank Height increase, March 27,1998 -
BRW-SE-1998-0065 Pressurizer Level Controller Setpoint Change, March 4,1998 l BRW-SE-1998-0203 Addition of Freeze Seal to Service Water inlet and/or Outlet for i
Cubicle Cooler, February 15,1998 BRW-SE-1998-0390_ Drawing Change to Delete Solenoid from Valves, i February 25,1998
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BRW-SE-1998-0419- , Reverse Exhaust Fan Direction While Supply Fan is Out of Service, March 2,1998 BRW-SE-1998-0475 Component Cooling Heat Exchanger Drain, Fill, and Vent, March 10,1998
' BRW-SE-1998-0535 ' M20-1-97-002C, insulate Vertical Surface of Unit 1 Condensate l _
Storage Tank, March 26,1998 l
BRW-SE-1998-0589 Spent Fuel Cooling Skimmers out of Service, March 27,1998 Change to Action / Work Request Processing Procedure,
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BRW-SE-1998-0672 April 2,1998 BRW-SE-1998-0683 New Containment Tendon Water Surveillance Procedure,
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April 2,1998 BRW-SE-1998-0915 Installation, inspection, and Removal of Scaffolding Procedures, June 1,1998 BRW-SE-1998-1139 Discrepancy Between UFSAR Figure and installed Configuration, June 12,1998 BRW-SE-1998-1140 Permit Heavy Load Travel along North Side of Spent Fuel Pool June 12,1998 i BRW-SE-1998-1141 Radwaste Filter Cubicle Shield Hatch Cover Sealant Material June 12,1998 1 BRW-SE-1998-1432 125Vdc Battery Discharge Duty Cycle, July 15,1998 BRW-SE-1998-1532 Auxiliary Feedwater Day Tank Sightglass Modification Test
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July 29,' 1998 BRW-SE-1998-1680 Demineralizer Resin Sluice Water Spool Piece Isolation Valves, August 10,1998
- . BRW-SE-1998-2169
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Belzona Application for Erosion / Corrosion Protection, ( _
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October 1,1998 ,
b BRW-SE-1999-0013 Drawing Change to Revise Valve Operator Position on Drawings, !
January 6,1999 ;
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BRW-SE-1999-0100 Drawing Revision to Show Valve Closed Rather than Open,
, January 26,1999 J
BRW-SE-1999-0101 Bypass Leakage Change from 0.05 Percent to 1.0 Percent, January 27,1999 10 CFR 50.59 Screenings BRW-FCS-1997-2027 Replacement of Emergency Diesel Generator instantaneous
' Preposition Board, January 20,1998 BRW-FCS-1998-0052 Rough Sample Cooler Pumps Piping Union installation, January 20,1998 BRW-FCS-1998-1580 Equipment Drain Line Drawing Changes, July 30,1998 BRW-FCS-1998-1687 - Chemical Feed System Administrative Out of Service, August 12,1998 BRW-FCS-1998-1864 Out of Service to Maintain Roll-up Doors Closed, August 11,1998 BRW-FCS-1998-1865 Unassigned Motor Control Center Cubicle Parts Removal, August 12,1998 BRW-FCS-1998-2052 Service Building Heating, Ventilation, and Air Conditioning System Cooling Coil Long Term Out of Service, September 18,1998 BRW-FCS-1998-2053 Cabinet Heater Long Term Out of Service, September 18,1998 BRW-PTES-1998-0338_ ASME Pump Surveillance for Control Room Chilled Water Pumps and Discharge Check Valves, February 20,1998 BRW-PTES-1998-0469 Eddy Current Testing Procedure Development, March 7,1998 )
BRW-PTES-1998-0471 Procedure Deletion and Replacement for improved Technical I Specifications (ITS), March 8,1998 BRW-PTES-1998-0485 Procedure Revisions to incorporate iTS into Surveillances, March 10,1998 BRW-PTES-1998-0637- Procedure Changes to BWVS Procedures, January 27,1998 .
- BRW-PTES-1998-1041 . Seismic / Dynamic Qualification Reports Review Procedure, ]
June 1,1998 BRW-PTES-1998-1227 : Replacement of Previous Screening Due to Procedure Revision, June 23,1998 BRW-PTES-1998-2004 New Procedure NEP 12-05," Control, Management and Review of Controlled Work,"" September 11,1998 BRW-PTES-1998-2264 Check Valve Monitoring and Preventive Maintenance Procedure, October 16,1998 BRW-PTES-1998-2342 Westinghouse Rod Control Maintenance Procedure, ,
October 23,1998 J BRW-PTES-1998-2650 Containment Tendon Water inspection Surveillance, ;
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December 3,1998 BRW-PTES-1999-0240 Shutdown Safety Management Program, February 23,1999 )
i Miscellaneous l Appendix R Emergency Lighting Summary of Survey Results ;
Appendix R Emergency Lighting Failure Summary, August 1998 to February 1999 ;
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EPRI TR-106826, Battery Performance Monitoring by Internal Ohmic Measurements l
Commercial Grade Survey of Woodward Governor Company, File 95-033, July 31,1995 I Maintenance Rule Performance Criteria for Appendix R Emergency Lighting System, Revision 0 Nuclear Design Information Transmittal NFM9800141 ECCS Pump Curves Used for Safety Analysis, July 15,1998 Nuclear Safety Review Board (NSRB) Charter, April 21,1998 NSRB Meeting Minutes 98-03, December 8,1998 NSRB Meeting Minutes 99-01, February 9,1999 NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants Quality Assurance Program Topical Report, Revision 66 Vectra Letter, " Summary of Observations and Lessons Leamed during the 2301 A Governing System Installation and Testing at Nebraska Public Power's Cooper Nuclear Station,"
March 15,1996 10 CFR 50.59 Summary Report June 19,1996 through June 18,1998
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