IR 05000266/1999005
| ML20205S710 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 04/20/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20205S707 | List: |
| References | |
| 50-266-99-05, 50-266-99-5, 50-301-99-05, 50-301-99-5, NUDOCS 9904260329 | |
| Download: ML20205S710 (41) | |
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! 1 U.S. NUCLEAR REGULATORY COMMISSION REGIONlil j l
Docket Nos: 50-266;50-301 l License Nos: DPR-24; DPR-27 i
Report Nos: 50-266/99005(DRS); 50-301/99005(DRS) l l
Licensee: Wisconsin Electric Power Company Facility: Point Beach Nuclear Plant, Units 1 & 2 Location: 6610 Nuclear Road l Two Rivers, WI 54241
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Dates: February 22 - March 12,1999 Team Members: A. Dunlop, Reactor Engineer, Team Leader C. Brown, Reactor Engineer K. Green-Bates, Reactor Engineer J. Neisler, Reactor Engineer G. O'Dwyer, Reactor Engineer D. Schrum, Reactor Engineer R. Winter, Reactor Engineer C. Baron, NRC Contractor Approved by: John M. Jacobson, Chief, Mechanical Engineering Branch Division of Reactor Safety 9904260329 990420 PDR G ADOCK 05000266 PDR
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EXECUTIVE SUMMARY Point Beach Nuclear Power Plant, Units 1 & 2 NRC Inspection Reports 50-266/99005; 50-301/99005 The NRC conducted an announced inspection to review the effectiveness of the engineering
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and technical support organization in the performance of routine and reactive site activities and the effectiveness of the licensee's controls in identifying, resolving, and preventing problem Enaineerina
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The methods used to control design changes and modifications were effectiv Modification packages were complete and of good technical quality. Plant changes were adequately designed and installed; and effective post-modification testing was specified and performed Design configuration and configuration controls were maintained l throughout the modification process. The licensee had effectively improved the procedures for design change control. (Section E1.2)
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In general, the 10 CFR 50.59 screenings and evaluations reviewed were appropriately prepared, of good quality, and were consistent with licensee procedures and regulatory requirements. The program ensured that trained and qualified personnel performed the necessary reviews and evaluations. Most evaluations adequately addressed the effects of the proposed changes on plant equipment and whether an unreviewed safety question existed. (Section E1.3)
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The additional oversight processes added to the safety review program demonstrated a rigorous effort and commitment to improve the thoroughness and consistency of 10 CFR 50.59 screenings and safety evaluations. (Section E1.3)
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The control of temporary modifications was good and the number of installed temporary modifications was not excessive. No concems were identified with temporary modification designs or their associated 10 CFR 50.59 screenings and safety -
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The current calculation process was effective in ensuring calculation adequacy, control, )
and compliance with regulatory requirements. The methods used in performing and 4 I
revising design calculations for recent modifications and design changes were found to be correct and appropriate. (Section E1.5)
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The engineering and operational actions taken in response to the higher than design auxiliary feedwater pump flow rates were appropriate to address the operability of the pumps during a loss of condensate storage tank event. (Section E1.5)
i The licensee had implemented an aggressive process to identify plant problems, which L
resulted in a large number of items in the engineering area. A work backlog review had been performed to ensure that all operability concems and outstanding corrective
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actions were addressed. The current engineering backlog assessment process provided an effective means of ensuring conditions adverse to quality were corrected in a time frame commensurate to their impact on plant safety. Recent actions to provide station personnel with management's expectations regarding timeliness of resolving issues, demonstrated the licensee's plans to address this issue. Based upon the engineering backlog trends, the efforts underway to reduce the backlog were progressing in a positive direction. (Section E1.6)
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The corrective action program was effectively implemented. Issues were being properly ;
identified, root cause evaluations were of good quality, and corrective actions were
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comprehensive. The large backlog cf 'ssues and timeliness of implementing corrective l actions and issuance of root cause evaluations reports was a concern. Hovaver, recent management incentives for backlog items were expected to resolve these fmeliness concerns. (Section E2.1)
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The systern engineers interviewed appeared to be qualified and knowledgeable of their )
systems and system issues even though some system engineers were new to the system or the plant. The system notebooks were a good tool to keep relevant information for a system in one place, however, there was a wide variance of quality in !
the system notebooks. (Section E2.2)
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Overall, the material condition and housekeeping of the systems reviewed were acceptable, and the ability of engineering personnel to identify material condition problems was good. The system engineers were knowledgeable about their system and familiar with their system's problem areas, backlogs, and modifications. (Section E2.3)
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The operability determination process was effective in ensuring that adequate bases were provided for operability determinations. However, there was a concern with the high number of operable, but degraded equipment requiring operability determinations and the length of time allowed to resolve the problem and restore the equipment to it's design basis. (Section E2.4)
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The industry Operating Experience Review Program was effective, based on the
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demonstrated implementation of plant changes resulting from the thorough screenings and detailed analyses of industry information. (Section E2.5)
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In general, procedures used in the engineering process were acceptable. A non-cited violation was issued for failure to perform independent verification of lifting and restoring wires. (Section E3.1)
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The quality assurance audits and surveillances were being performed by qualified personnel. The results of the audits and surveillances were in-depth and identified a number of significant findings. The corrective actions for findings of the earlier audits resulted in program improvements. (Section E7.1)
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The plant review organizations provided the technical expertise in the multiple areas required to review and evaluate the documents and activities that occur at the plan Members of the groups were experienced and were aggressive in pursuing plant problems and issues. The minutes of the meetings indicated that probing questions were asked. The plant review organizations provided an effective and vigorous overview of plant activities related to safety. (Section E7.2)
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Report Details l 111. Enaineerina The review of engineering included both design and support engineering activities. The design ,
review included design changes, temporary modifications, and 10 CFR 59.59 evaluations and l l screenings. Engineering support included system engineering, problem resolution and l corrective action activities, as well as normal engineering involvement with operations, maintenance, and other plant organizations.
i E1 Conduct of Engineering E General Observations on the Enaineerina Oraanization (37550)
? l The conduct of engineering at Point Beach had undergone a number of changes during I the past year and was still in a state of transition at the time of the inspection. This change was a result, in part, of a number of significant programmatic and process issues that were identified by the quality assurance (QA) department during the 19% - 1998 time period. In order to resolve some of these issues, the licensee was in the process of reorganizing the engineering department. This included incorporating outside personnel as engineering managers into the organization, increasing the size of the engineering staff, and moving design engineering from the corporate office to the site. A Performance improvement Plan was established to prioritize, schedule, and address the major programmatic issues within the engineering organization. Several of these areas were reviewed during this inspection including the modification process,10 CFR 50.59 safety evaluations, equipment material condition, system engineering, and prioritization and management of work. While some changes had resulted in positive program improvements, changes in other areas were not yet complete or in place for a sufficient length of time to evaluate their effectiveness during this inspectio E1.2 Rasian Chances and Modifications (37550) l Insoection Scooe The methods used to control design changes and modifications were reviewed to verify adequacy, control, and compliance with regulatory requirements. The review included the basic design change procedures and 21 selected modification packages. The modification packages selected included modifications installed within the previous 2 years. Modification packages were reviewed in detail, discussed with cognizant system and design engineers, and the inspectors walked down accessible portions of some modification insta!lations. Elements of this review included 10 CFR 50.59 evaluations, screenings, and design calculations, which are discussed in other sections of this report.
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l Observations and Findinas
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l Overall the packages reviewed were complete, of good quality and were adequate to accomplish the design changes. The team noted a few minor administrative errors that were corrected by licensee personnel. For example, the reference number of a safety i
screening for a modification package had been omitted. Calculations supporting the
- design changes reviewed were of good technical quality and supported the design changes. Walkdowns indicated that modifications were properly installe Post-modification testing was effective in verifying that the modified system would perform its designed function. A sample of affected drawings were verified to be revised or red-lined as appropriate and, when necessary, operating procedures were revised or written. Two concerns identified during the . modification package review were the l licensee's independent verification program associated with the wiring in the main control
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boards modification as discussed in Section E3.1, and concerns with the safety evaluation for a fire damper modification as discussed in Section E Prior to this inspection, licensee management had recognized a need to improve the design change procedures and had made numerous revisions. The team determined that the procedures were effective as no procedure-caused problems were identified in the reviewed design modifications. On January 1,1999, design authority and control of l design work were transferred to the site from the corporate offices in Milwaukee. The
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team detected no instances where the transition had caused problems, although all the modifications reviewed during this inspection had been done while design authority was in the corporate office - Selected site and corporate design personnel were interviewed by the team and found to be knowledgeable of assigned engineering tasks. As discussed in Section E1.6, the team concluded that the licencee were taking appropriate measures to reduce the large modification backlo Conclusions The team concluded that the methods used to control design changes and modifications were effective. Modification packages were complete and of good technical qualit Plant changes were adequately designed and installed; and effective post-modification testing was specified and performed. Design configuration and configuration controls were maintained throughout the modification process. The licensee had effectively improved the procedures for design change contro j
E CFR 50.59 Evaluations and Screeninas (37001)
a.- Inspection Scope
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The methods and procedures used to control 10 CFR 50.59 screenings and evaluations
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were reviewed to verify adequacy,' control, and compliance with regulatory requirements.
i Emphasis during this review was on design changes and modifications. The
- 10 CFR 50.59 screenings and evaluations were discussed with cognizant licensee I- personnel and selected evaluations were reviewed in detail to verify implementation and compliance with the requirements of 10 CFR 50.5 I
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10 CFR 50.59 Safety Evaluation Procedure Recent changes to NP 10.3.1, " Authorization of Changes, Test, and Experiments (10 CFR 50.59 and 72.48 Reviews)," resulted in improved procedura guidance for screenings and performing 10 CFR 50.59 safety evaluations (SEs). The procedure effectively assigned responsibility for key areas to assure that 50.59 SEs were effectively prepared, reviewed, and approved. The procedure contained a comprehensive listing of licensing basis documents, although it did not list design basis documents that were used to prepare 50.59 SEs. In addition, the procedure listed other documents that may contain regulatory commitments to facilitate preparation of 50.59 SEs. Adequate procedural guidance existed for maintaining records and formally reporting to the NRC the changes, tests, and experiments made in accordance with 10 CFR 50.5 CFR 50.59 Safety Reviews for Plant Modifications in general, the screenings and SEs reviewed were of good quality. The screenings and SEs reviewed, were very detailed as to the sections of the Technical Specification (TS)
and the Final Safety Analysis Report (FSAR), which could be construed as being relevant to the change. Evaluations were performed for plant changes when required, and the evaluations were generally thorough with 10 CFR 50.59 questions appropriately answered. The team did not identify any screenings where SEs were required and had not been performed; nor did the team find any unidentified unreviewed safety questions (USO). Sufficient detail was typically provided so that an independent reviewer could also arrive at the conclusion that no USQ existe Although most of the evaluations reviewed adequately assessed how the physical change would affect the structure, system, or component described in the applicable FSAR section, TS, or associated dependent document; the team did identify two examples where the 50.59 process was not comprehensive:
Modification 97112*A. "G-01/G-02 Fire Damner Modification:" This modification was performed to prevent unintended closure of fire dampers used in the emergency diesel generator (EDG) room ventilation system, because that event would render the EDGs inoperable. The switches, wiring, and solenoids that closed the fire dampers after an 18-second delay via a manual trip function were removed. As fans and dampers could no longer be manually tripped, fan control was solely based on room temperature, such that a fire in the EDG room would cause the ventilation fans to start and continue operating until temperature in the room reduced below the trip setpoin The team noted that the SE did not appear to sufficiently evaluate all aspects of the modification such as automatic fire damper operation, seismic qualifications, and high energy line break issues. The licensee subsequently provided additional documentation that adequately addressed the seismic qualifications and high energy line break l concerns.
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With respect to the automatic fire damper operation issue, the SE assumed that the heat from the fire would melt the eight fusible links and isolate the EDG room from the turbine building. However, the SE did not appear to provide sufficient justification that the fusible l
links would melt as designed. The SE also stated that even if the fusible links did melt, the licensee believed that the dampers would not close against the air flow from the fan such that the fire dampers' automatic trip function may not work as designed. Therefore, the licensee included in its fire attack plan, instructions for the fire brigade to trip the fans in the EDG room, allowing the damperr to drop. If the entry into the EDG room was not possible due to the fire, the fans wouht be tripped at the fan breakers. The following issues identified during the inspection cequire follow up during a subsequent inspectio .
The EDG room fire dampers' operation would not be fully automatic as per design because the air pressure from the fans would not allow the dampers to clos .
The SE did not appear to provide adequate justification that the heat from the fire would melt the eight fusible links and isolate the EDG room from the turbine building. In addition, higher temperature rated fusible links were scheduled for installation as part of a high energy line break analysis, which would require further justificatio i
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The removal of the manual trip function in combination with the dampers potentially not automatically isolating, could result in a significant amount of additional time that smoke and heat would be exhausted into the turbine building before the fan breaker could be tripped. It was also noted that it would take a significant amount of additional time to manually close each dampe These issues are considered an Unresolved item (URI 50-266/301-99005-01(DRS))
pending further NRC review of the design requirements for the EDG room fire damper Condition Report (CR) 97-1822. "Imoroper Brass Fittinas in Waste Gas System:" This CR identified that brass material was used in waste gas system piping. The team noted that this was in conflict with the system material described in the FSAR (Chapter 11, !
Section 2.2), which stated that "all waste gas piping and all valves exposed to gases are j carbon steel." Although the FSAR system description was in conflict with the material installed in the waste gas system, the CR was closed out to work orders without being fully evaluated. Subsequent investigation by the licensee determined that the brass valves had been installed by a 1977 modification and that the valves did not meet the operability requirement described in TS 15.1.C. The licensee issued CR 99-0767 and performed an operability determination. The licensee determined the discrepancy was not a safety issue, which was considered acceptable to the tea Utilization of Current Licensina Basis in 10 CFR 50.59 Safety Evaluations
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The team reviewed the processes that ensured the design information necessary for
- preparing adequate 10 CFR 50.59 SEs were available to the personnel that prepared ,
and reviewed SEs. The team determined that the current licensing basis was available for review as part of the licensee's new, easily accessed electronic database. In
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addition, in response to a QA audit finding concerning lack of training, all employee's
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computer training class to improve knowledge and performance when searching the current licensing basi The team noted that in response to previously identified problems with the thoroughness and consistency of 10 CFR 50.59 screenings and SEs, licensee personnel had developed and implemented a number of actions and oversight processes. This included: revising site procedures to clarify safety review expectations and provide stronger links with the 10 CFR 50.59 program; issue a 50.59/72.48 Feedback Newsletter l
to provide staff feedback and suggestions; establish the Point Beach Safety Evaluation Steering Committee to provide 50.59 process oversight; and assign new 50.59 process
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review responsibilities to the Managers Supervisory Staff and the 10 CFR 50.59 Review l Subcommittee.
l l Conclusions In general, the 10 CFR 50.59 screenings and SEs reviewed were appropriately prepared, of good quality, and were consistent with licensee procedures and regulatory i requirements. The program ensured that trained and qualified personnel performed the l necessary reviews and SE preparation. Most SEs adequately addressed the effects of l the proposed changes on plant equipment and whether an USQ existe The team concluded that the additional oversight processes added to the safety review program demonstrated a rigorous licensee effort and commitment to improve the thoroughness and consistency of 10 CFR 50.59 screenings and SE E1.4 Temoorary Modifications (37550) I insoection Scope The methods used to control temporary modifications were reviewed to verify adequacy, control, and compliance with regulatory requirements. The review included controlling procedure NP 7.3.1, " Temporary Modifications," and ten open temporary modification packages, which also included the appropriate 10 CFR 50.59 evaluations or screening Temporary modifications were discussed with cognizant licensee personnel and the team walked down accessible portions of installed temporary modification l Observations and Findinas i l
l The team did not identify any concerns with the control of temporary modifications. The temporary modification process provided adequate controls for temoorarily installed l material and equipment. There was adequate engineering involvement and control of
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l The procedure stated that temporary modifications not removed during a unit's next refueling outage (or 6 months if not outage related) must have a justification to remain installed. Of the 25 open temporary modifications, only 2 had been installed longer than 1 year. Appropriate justifications had been initiated for these temporary modifications.
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The number of temporary modifications was substantially reduced during the recent l
refueling outag l The licensee had corrected the majority of the temporary modification problems identified during the May 1998 QA audit report on design engineering. Additional corrective
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modifications performed in December 199 Conclusions The team concluded that control of temporary modifications was good and the number of installed temporary modifications was not excessive. No concerns were identified with l temporary modification designs or 10 CFR 50.59 screenings and SEs. Engineering personnel were adequately involved in the temporary modification proces E1.5 Calculations (37550) Inspection Scope The methods used in performing and revising design calculations for modifications, design changes, and setpoint changes were reviewed to verify adequacy, control, and compliance with regulatory requirements. The review included the controlling procedure and 23 selected design calculations, which supported recent modification packages, setpoint changes, and operability determinations. The calculations were reviewed for accuracy and to verify appropriate inputs, assumptions, and calculation methods. The calculations were discussed with cognizant licensee personne Observations and Findinas Overall, the team determined that the methods used in performing and revising design calculations were correct and apprapriate. The calculation purpose, assumptions, method, references, and inputs were clearly stated. Calculation revisions were clearly 1 identified. The team noted DG-101," Design Guideline," was used as the method and j basis for instrument channel setpoint calculation preparation, review, approval, and loop !
accuracy analysis. The team determined the methodology was technically sound, consistent with current industry practices, and sufficiently comprehensive to evaluate indication uncertainties. Discussions with licensee engineering personnelindicated that engineering assurance reviews have been implemented for a sample of the calculations completed each month as part of the licensee's engineering Performance improvement Plan.
l The team identified one condition where a design calculation for a safety-related system contained non-conservative methodology. Calculation 97-0172,"Available Water in
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Volume of Piping to the Auxiliary Feedwater (AFW) Pumps Following Pipe Break at Elevation 25-6," was performed to determine the volume of water that would be available in the common AFW pump suction piping in the event of a tornado or seismic event
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resulting in the loss of the condensate storage tanks. Calculation 97-0172 provided input to calculation 97-0215, " Water Volume Swept by all Four AFW Pumps Following a Seismic / Tornado Event Affecting Both Units." Calculation 97-0215 concluded that the AFW pump low suction pressure trip feature would be adequate to protect the AFW pumps from damage and that the suction volume that would be consumed by the pumps until the pumps tripped would not exceed the available volume in the protected suction j pipin Calculation 97-0172, however, used a non-conservative methodology in determining the l water volume available in the horizontal portion of the common AFW pump suction header. The calculation determined that the full volume of water located above the bottom of the branch connections to the individual AFW pumps would be available to all the operating pumps. This calculation did not address the potential of air entrainment to the operating pumps when the branch connections partially uncovered. Significant air entrainment could potentially cause all the AFW pumps to become unavailabl As a result of this concern, the licensee initiated CR 99-0795 and associated operability determination (OD). The OD evaluated the potential of air entrainment and determined that the motor-driven AFW pumps would trip about the same time that their suction lines would transition to open channel flow, and that the turbine-driven pump suction lines would transition to open channel flow about 3 seconds before the turbine-driven pumps would trip. Based on the configurations of the individual AFW pump suction lines, the l OD conclusion stated that the postulated air entrainment mechanism would not jeopardize the operability of the AFW pumps before, during, or after the pump suctions were shifted to the service water system and the pumps restarted. Licensee personnel stated that calculation 97-0172 would be revised to address this issue. The team concluded the licensee's justification was appropriate to address the concer During the investigation of this issue, the licensee discovered that the Unit 2 turbine-driven AFW pump discharge valves,2AF-4000 and 2AF-4001, had been incorrectly positioned during the February 28,1999, performance of IT 09A, " Cold Start of Turbine Driven Auxiliary Feed Pump and Valve Test Unit 2 (Quarterly)." The valves had been adjusted to allow a flow approximately 100 gallons per minute above design flow rates. The licensee issued CR 99-0801 to address this condition. As a result, the licensee declared all four AFW pumps inoperable and entered TS Limiting Condition for ,
Operation action statement 15.3.0.B. The 1:censee subsequently tripped the Unit 2 l turbine-driven AFW pump, entered TS Limiting Condition for Operation action statement i 15.3.4.C.1, declared the remaining three AFW pumps operable, and exited action statement 15.3.0.B. This event was reported to the NRC on March 12,1999, in j accordance with 10 CFR 50.72(b)(1)(ii)(B). The licensee successfully performed IT 09A to restore valves 2AF-4000 and 2AF-4001 to their correct positions. The Unit 2 turbine-driven AFW pump was then declared operabl s a Conclusions The team concluded that the current calculation process was effeciive in ensuring q calculation adequacy, control, and compliance with regulatory requirements. The i methods used in performing and revising design calculations for recent modifications and design changes were found to be correct and appropriat The engineering and operational actions taken in response to the higher than design AFW pump flow rates were appropriate to address the operability of the pumps during a loss of condensate storage tank even E1.6 Enaineerina Backloa Inspection Scope l
The methods used in evaluating, prioritizing, scheduling, monitoring, and resolving j engineering issues were reviewed to verify that adequate processes were in place to properly manage the engineering backlog. The review included the controlling processes established to prioritize issues and a review of the selected system's engineering backlog for adequate imple nentatio Observations and Findinas l
The team noted that licensee mant.gement had lowered reporting thresholds in 1997 and j 1998 to address long-standing equipment and process issues, which resulted in a dramatic increase in problem identification CRs. As approximately 50 percent of the corrective actions were associated with engineering, a large engineering backlog quickly materialized. Due to the large backlog, a process was established in 1998 to review the backlog, establish an improved priority system based on a simplified 1 (high safety significant) to 5 (low significant) strategy, and to plan for reducirig the backlog. In l
October of 1998, licensee management established new expectations to focus on higher
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priority 1 and 2 issues with increased accountability for all management and staff personne The team obser/ed that recent progress had been made towards decreasing the overall backlog. Several significant engineering issues had been resolved, with improvements implemented or well-underway. These issues were in the areas of: 10 CFR 50.59 SEs, core design, Appendix J, maintenance rule, inservice test, and relief valves. In addition, the 1996 probabilistic safety assessment was updated and incorporated into the maintenance rule. The team also noted that probabilistic safety assessment results were being used in plant modification decision However, a number of significant issues remained to be resolved such as: the Appendix R rebaseline, American Society of Mechanical Engineers (ASME)Section XI inservice inspection program, and drawing controlissues. In addition, many significant equipment issues needed to be addressed. At the time of the inspection, there were 55 plant upgrade projects and process or program improvements within the engineering
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Performance Improvement Plan. The team noted that the root cause evaluation and preventative maintenance improvement projects were behind schedule, but the remainder of the projects appeared to be on time or ahead of schedule. An Engineering Advisory Committee had been established in January 1999 to provide oversight and monitoring of the engineering backlog and any proposed new items. In an attempt to 1 manage allocation resources more effectively such that priority issues could be addressed at a faster rate, items within the backlog with the lowest priority were currently 4 I
under critical review by the committee along with input from plant engineers, probabilistic safety assessment experts, and plant personal, to determine possible cancellation of low priority items. This project was expected to further reduce backlog without reducing j safety, Conclusions
The team concluded that the licensee had implemented an aggressive process to l identify plant problems, which resulted in a large number of items in the engineering l system. A work backlog review had been performed to ensure that all operability concems and outstanding corrective actions were addressed. The current engineering backlog assessment process provided an effective means of ensuring conditions adverse to quality were corrected in a time frame commensurate to their impact on plant safety. Recent actions to provide station personnel with management's expectations regarding timeliness of resolving issues, demonstrated the licensee's plans to address this issue. Based upon the engineering backlog trends, the efforts underway to reduce the backlog were progressing in a positive directio E2 Engineering Support of Facilities and Equipment The team reviewed the effectiveness of engineering support to plant organizations, which included plant management, operations, and maintenance. Much of the engineering support involved assistance in the documentation, evaluation, and resolution of problem E2.1 Corrective AdLon Proaram (40500) Inspection Scoce The methods used to control the corrective action process at Point Beach were reviewed to verify the adequacy and effectiveness of identifying and correcting plant problem The review included applicable controlling procedures, records, and internal audits, as well as an in-depth review of selected condition reports (CRs).
Timeliness and priority of actions completed or scheduled were considered, as well as the tracking of actions to correct or minimize problems. Discussions of corrective issues were held with station personnel in the operations, maintenance, and engineering departments. Root cause evaluations (RCEs) were evaluated for initial problem identification and characterization, assessment of operability, immediate corrective actions, corrective actions to prevent recurrence, and evaluation of repetitive problem l s
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Condition Reports The licensee's corrective action program used CRs to identify and correct problem Quality condition reports (OCRs) were CRs that the quality group initiated and close The initiator of a CR/QCR described the problem and indicated its significance level from .
high to low (A, B, C, or D). The CR was screened for operability and reviewed at the I next daily morning meeting. Each corrective action assigned to the CR, was divided into individual tasks (denoted by an action number) to be performed within the corrective action process. Each action number was given a priority and a due date. The significance of a condition or event that would initiate a CR was determined to have a low 4 threshold (i.e., if there was question about writing a CR, then write a CR). When all the corrective actions for a CR were complete, the designated punchlist administrator reviewed the actions prior to closing the C The team noted that the number of new CRs due to a lowered threshold had resulted in a large backlog, and that timeliness for completion of the CRs had not been good. The licensee's internal evaluation of CRs revealed a large number of items overdue by as much as 6 months, but that resolution of level A and B CRs were generally more timel Recent actions to provide station personnel with management's expectations regarding timeliness of resolving issues, demonstrated plans to address this issu Root Cause Evaluations The RCEs provided an evaluation of A and B level CRs to identify the root cause of an event in order to prevent recurrence. The RCEs reviewed were generally of good quality. The basic cause of an event appeared to be identified for each occurrenc !
Barrier Analysis and Event and Causal Factor Chart were two of the methods used to perform RCEs. Barrier Analysis was used to identify barriers that should have prevented the event. The evaluation indicated the failure, or absence, of a preventative measure, resulting in the development of an effective barrier. The Event and Causal Factor Chart method was used to graphically present the events and operations that preceded a failure event. Chart evaluation then demonstrated where failures in operation of equipment, ineffective procedures, etc., contributed to the event. The results of the evaluation indicated areas where corrective actions would be needed to preclude reoccurrenc Several RCEs included a historical list of similar events to determine possible repetitiveness. The team also noted that the issuance of the RCE document was not always timely, generally 4 to 8 months after the event. As previously stated, the recently issued management expectations on timeliness and accountability would be expected to improve issuance of RCEs.
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, Conclusions The team concluded that in general, the corrective action program was effectively implemented. Issues were being properly identified, RCEs were of good quality, and corrective actions were comprehensive. The team was concerned with the backlog of issues due to timeliness of implementing corrective actions and issuance of RCE report However, recent management incentives for backlog items were expected to resolve these timeliness concern E2.2 System Enaineerina (37550) Inspection Scope The team interviewed system engineers and engineering supervisors and reviewed l system notebooks for the service water (SW), auxiliary feedwater (AFW), chemical and
! volume control system (CVCS), EDG,4160 volts alternating current (VAC), and 480 VAC l systems. The team also reviewed the licensee's self-assessment on system
! engineering, SA 98-16, " Point Beach System Engineering Self Assessment for l July 1998."
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l Observations and Findinas
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The licensee had performed a vigorous self-assessment cf the system engineering l program in July 1998. The self-assessment documented system engineering strengths l as effectively supporting plant operations and the use of the System Engineering Review Board, which was found to be effective in identifying and resolving unit restart issues.
l The assessment also identified several weaknesses, including poor integration of new engineers and new work processes. This resulted in ineffective communication and
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reduced work productivity and, due to large workloads, system engineers were not meeting the management expectations expressed in SEM 1.0, " System Engineering i Handbook." The team noted that the licensee had planned remedial actions for the identified weaknesses, i
i Overall, improvements in system engineering were appropriate for the length of time the remedial actions had been in place. For example, additional engineers had been brought into system engineering to decrease the workload and more staff increases
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were planned. The team noted that some of the remedial actions for the
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self-ajsessment had not yet been completely implemented, e.g., the management expectations of the system engineering handbook for system walkdown documentation and system turnovers had not been fully met because of large workloads. The team also noted that system engineers did not appear to fully understand the importance of meeting the management expectations of the system engineering handbook. However, licensee management was already aware of this issue and had taken steps to convey to the system engineers the proper emphasis on meeting the expectation The system handbooks contained system descriptions, FSAR sections, design basis documents, system issues, vendor contacts, and other relevant material. The team 15-Li
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, i determined that the handbooks had been improved, however, there was a wide variance l of quality in the system notebooks. The team noted that while SEM 1.0 did not {
recommend that the system handbooks contain the TS sections applicable to the system, the system engineers had added the system-related TS sections to their handbooks for all the handbooks reviewed. The licensee was considering amending ,
SEM 1.0 guidance to include appropriate TS sections as a quality improvement ite ! Conclusions The team concluded that the system engineers interviewed appeared to be qualified and knowledgeable of their systems and system issues even though some system engineers were new to the system or the plant. The system notebooks were a good tool to keep relevant information for a system in one place, however, there was a wide variance of quality in the system notebook E2.3 System Walkdown Observations (37550) Insoection Scoce l
The team conducted walkdowns of the accessible portions of the SW, AFW, CVCS, EDG,4160 VAC, and 480 W C systems with the applicable system engineer to assess material condition, housekeeping, and the ability of engineering personnel to identify problem Observations and Findinas The walkdowns generally did not identify issues that had not been previously identifie Numerous deficiency tags were noted on plant equipment during the walkdowns indicating the licensee was properly identifying issues. In general, the team concluded I that the material condition and housekeeping was acceptable, although large system work backlogs indicated that plant material condition could be improved. The team determined that the system engineers were knowledgeable aoout the system, familiar with system problems, work backlogs, current modifications and the long-range plan for their system. Specific observations noted during the walkdowns were as follows:
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The team identified two pipe supports in the AFW pump room that had gaps between the baseplate and concrete wall that exceeded the criteria specified in procedure NDE 754, " Visual Examination (VT-3) of Nuclear Power Plant Components." The procedure had been recently revised to include an acceptance criteria for gaps, however, the baseplates in question had not yet been inspected to the newly revised procedure, and the team did not consider this matter a regulatory issue at this time. The licensee issued CR 99-0797 and performed an operability determination to evaluate the supports capability to adequately transfer the design loads to the building structure. The team concluded that the licensee was appropriately evaluating and prioritizing the identified support gaps.
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The material condition of the two electrical systems, including breakers, relays, indicators, cubicles, and cable trays, were acceptable except for a minor hardware problem on one breaker cubicle doo .
The CVCS system was generally in good material condition, and there appeared to be an improved control of boric acid leaks and fewer radiological areas when compared with conditions observed during past inspection Conclusions -
The team concluded that overall, the material condition and housekeeping of the systems reviewed were acceptable, and that the ability of engineering personnel to identify material condition problems was good. The system engineers were knowledgeable about their system and familiar with their system's problem areas, backlogs, and modification E2.4 Operability Determination Review (37550) Inspection Scooe The method used to control operability determinations (ODs) was reviewed to verify adequacy, control, and compliance with regulatory requirements. The review included 17 ODs associated with operable but degraded equipment. The determinations were reviewed for adequate Justification to ensure the system or component remained operable and corrective actions to restore the system or component to design specifications were adequate and timely, Observations and Findinas Overall, the team determined that the ODs documented an adequate basis for the conclusions. The ODs were found to provide appropriate references to engineering analyses and to identify operating limitations when required. The associated CRs were found to include the required actions to resolve the issues. However, the team was concemed with the high number of operable, but degraded equipment (~130) and the length of time allowed to resolve the problem and restore the equipment to design basi Due dates were routinely extended for resolution of issues, although recently issued licensee expectations addressed personnel accountability and the timeliness concern with resolving issues. In addition, the team was concerned with the number of ODs issued to address SW system components, particularly during periods of high lake water temperatures. Engineering personnel stated that ongoing efforts to restore the analytical design basis of the SW system will eliminate the requirement for these OD The team identified one condition where an OD for the SW system included an inappropriate justification for operability. This OD associated with CR 98-2704 was
- issued last summer to allow plant operation with SW inlet temperatures of up to 78' The previous analyses were based on a maximum temperature of 75'F. The OD stated,
"The CCW [ component cooling water) system is not assumed to be operable post-DBA
- (design basis accident) per OD 98-0564. Therefore, the effect of the SW temperatures on the CCW system is not evaluated."
The OD associated with CR 98-0564 was issued to address the potential of a CCW system leak inside containment as a result of a loss-of-coolant-accident. The OD stated that the probability of this event occurring would be on the order of 10~ 2 per reactor year of operation. The OD addressed mitigating this loss-of-coolant-accident event after the postulated CCW system failure and determined that the CCW system would be accessible for repairs during this event. This OD also included a reference to the TS requirements for operability of the CCW syste The team determined that the use of OD 98-0564 to support plant operation with SW inlet temperatures greater than the analyzed value without evaluating the effects on the CCW system was inappropriate. The licensee's engineering personnel initiated CR 99-0777 to address this issue and issued Revision 2 of the OD for CR 98-2704 during the inspection. Revision 2 of the OD appropriately addressed the effects of the elevated SW temperature on the CCW syste Conclusions The team concluded that, overall, the licensee's OD process was effective in ensuring that adequate bases were provided for ODs. However, the team was concemed with the high number of operable, but degraded equipment requiring ODs and the length of time allowed to resolve the problem and restore the equipment to design basi E2.5 - Industry Ooeratina Exoerience Review Prooram (37550. 40500) l Inspection Scope l
The methods to use and control outside or industry information and problem notifications j such as generic letters, information notices, Institute of Nuclear Power Operations notifications, and service information letters, were reviewed to verify adequate evaluatbn for plant applicability and subsequent action. The review included the controlling procedure and the actions taken on selected notification documents as well as discussions with Industry Operating Experience (OE) Review Program personne Observations and Findinas The team observed that OE was used to assess industry information for events which impacted nuclear safety issues. The pnmary objective for each assessment was to provide data that could be used for problem solving and preventive measures, and to j
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transfer the lessons leamed into corrective actions that would prevent a similar occurrence at Point Beach. The health of the program had been addressed by an !
internal bench marking effort against industry standard institute of Nuclear Power l Operations97-011, " Guideline for the' Use of Industry OE" in June of 199 p
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As a result of the self-audit, changes to NP 5.3.2 " Industry Operating Experience Review Program," were made that resulted in improved procedural guidance for program trending, Nuclear Tracking System prioritization, and incorporation of program ' lessons i learned'into training programs. The procedural guidance clearly defined program l responsibilities for the dissemination, screening, investigation, evaluation, disposition, l scope, and program effectiveness reviews. The team reviewed a number of extemal l industry operating experience issues and the licensee's assessment and corrective actions associated with the issues. The issues reviewed were handled in accordance l with procedural guidance and appeared to be appropriately dispositioned. Examples where the OE program initiated plant changes included:
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OE 9587: 4160 Volt Breaker Anomaly - The licensee identified that the same issue with the Westinghouse 50-DH-350 breakers existed at Point Beach. As a result, CR 99-0251 was issued and a Notice of Enforcement Discretion was obtained from the NRC to allow the unit to operate while repairs / modifications were made to the breaker OE 8942: Alarms Disabled By improper Recorder Operation - The licensee identified that although Point Beach did not use any Yokogawa recorders of the subject model, the model Yokogawa recorders used at Point Beach had the i same design weaknes *
OE 8714: Inadvertent Dilution During Low Power Operation at Vogtle - The licensee added precaution and limitation notes to applicable operating l
procedures and operating instructions to prevent a similar occurrenc Conclusions i
The team concluded that the Industry Operating Experience Review Program was ;
effective, based on the demonstrated implementation of plant changes resulting from the i thorough screenings and detailed analyses of industry informatio l E3 Engineering Procedures and Documentation E3.1 Enaineerina Procedure Review (37750) Insoection Scope The team reviewed the procedure guidance controlling the activities observed or examined by the team. In addition; engineering managers and workers were interviewed to assess their knowledge of procedure conten Observations and Findinas in general, procedures reviewed during the inspection provided sufficient guidance to perform the tasks in question. Numerous procedures had been recently revised to resolve concerns identified during QA audits and self-assessments to improve
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engineering processes. Specific concerns identified with procedural guidance are discussed below, b1 Indeoendent Verification During the review of documentation from the main control boara wiring separation modifications93-056 and 97-007, the team noted inconsistencies in the documentation l of independent verification of lifted leads. Specifically, Form PBF-0036, " Wire Removal
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Log," had requirements for personnel to affix their initials indicating a first person
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checker, wire remover, wire restorer, and second person checker. On 9 of 20 forms !
I reviewed by the team, checker initials and individual performing the work were the same {
person. Procedure NP 2.1.2, " Independent Verification and Concurrent Checks," stated 1 an independent verifier was a second person not involved with the task and that independent verification was required for lifting and restoring wires in safety system l
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10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings," required )
l activities affecting quality shall be prescribed and accomplished by documented instructions, procedures, and drawings. Contrary to the above, nine examples were identified where independent verification was not performed for lifting and restoring wires associated with the main control board wiring modification in accordance with NP 2. This Severity Level IV violation is being treated as a Non-Cited Violation, consistent with )
Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective l action program as CR 99-0750 (NCV 50-266/301-99005-02(DRS)).
b.2 Safetv-Related Pioe Suocort VT-3 Inservice Inspection Procedure As discussed in Section E2.3 of this report, the team identified two AFW pipe supports in the AFW pump room with gaps between the baseplate and concrete wall which exceedM the criteria specified in procedure NDE 754. Procedure NDE 754 established formal guidance for gap criteria, but did not include guidance on the disposition of gaps that exceeded the criteria. The procedure was unclear from the standpoint of assuring that the appropriate plant departments were contacted when pipe support gap criteria l was exceeded. The licensee issued CR 99-0766 to revise the procedure to provide guidance for appropriately disposition gaps that exceed the approved criteri b.3 Emeraency Liahtina Surveillances The team reviewed procedure MWP 126," Emergency Lighting Performance Monitoring Test," which had been recently revised to implement impedance testing. The team noted that a quarterly surveillance test PBP E-M3-1, " Emergency Lighting," was a prerequisite to the impedance test. However, it did not appear the quarterly surveillance tests were {
! being performed prior to the impedance test as required. The maintenance staff stated the PBP E-M3-1 instruction sheet was used as the prerequisite to the impedance test, '
but test results were not documented. This was considered a weakness in the use of this procedure.
l The team identified additional problems with call-up activity E-A2 and surveillance procedure MWP 126. For example, Step F(1) of E-A2 called for adjusting or replacing
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the charger module if the battery voltage dropped below 6 volts. However, the correct battery range should have been 7.2 volts +/ .3 volts as stated in PBP E-M3-1. The licensee stated the problem occurred because the call-up was performed prior to the call-up being revised to include the correct battery range. The incorrect battery range did not effect the actual work performed on the battery. Procedure MWP 126 did not include a temperature correction calculation if the battery temperature was outside of the temperature range specified in the procedure as recommended by the Electric Power Research Institute (EPRI) document TR-106826 " Battery Performance Monitoring by Internal Ohmic Measurements." Subsequent to the inspection, on March 23,1999, the licensee obtained a memorandum from a member of the EPRI research team that stated temperature compensation was no longer recommended because it was a small variable in relation to other variables that influence the actual measured ohmic value. In addition, the MWP 126 did not state the purpose of the charger voltage data or provide acceptance criteria for the data. The licensee initiated CR 99-0772 to address the team's issue Conclusions in general, procedures used in the engineering process were acceptable. A non-cited violation was issued for failure to perform independent verification of lifting and landing electrical leads as required by procedure NP 2. E7 Quality Assurance in Engineering Activities Quality assurance of engineering activities was provided by the QA organization, who performed surveillances, assessments, and required audits of plant activities, in addition, several committees provided an independent review of selected documents, problems, and issues. These activities provided substantial support to corrective action and problem correction and avoidance type activitie E7.1 Audits and Surveillances (40500)
a. Inspection Scope The methods used to perform and control the QA (now quality verification) audits and surveillances were reviewed to verify adequacy, control, and compliance with regulatory requirements. The review included the controlling procedures and selected 1997,1998, and 1999 audit and surveillance records as well as discussions with cognizant licensee personnel, b. Observations and Findinas Audits were scheduled and routinely a planned function of quality verification. The audits covered a broad scope, usually a program. The surveillances were usually narrow in scope and less formally planned. The audit plan and checklists were contained in the audit result documents. Personnel performing the audits and surveillances were well-qualified, experienced, and cognizant of the audit or surveillance areas. If a
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l technical specialist was used on the evaluation team, the specialist's resume was l included in the results documentation. Recent audits performed included:
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Inservice inspection and Repair / Replacement Programs )
Ccrrective Action and Operating Experience Programs
10 CFR 50.59 Screening and Evaluation The audits and surveillances were in-depth and identified a number of significant findings that resulted in a concern with engineering effectiveness and the initiation of the Performance improvement Plan. A quality condition report (QCR) was issued for each finding to provide documentation and tracking. Documentation was of good quality and issued in a timely manner. The findings were of sufficient detail to provide adequate l Information for any necessary corrective actions. The quality group verified corrective actions were appropriate prior to close-out of each QC ! Conclusions l
The team concluded that QA audits and surveillances were being performed by qualified personnel. The results of the audits and surveillances were in-depth and identified a l number of significant findings. The corrective actions for findings of the earlier audits resulted in improvements in some programs as documented in other sections of this i repor l E7.2 Review Committee Activities (40500) .
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i The methods used by two separate and independent review groups were reviewed by the team to verify adequacy, control, and compliance with regulatory requirements. The review included the controlling procedures, meeting minutes, and other selected records of activities. The groups reviewed were the Off-Site Review Committee (OSRC) and the Manager's Supervisory Staff. The functions, findings, and activities of the two groups were discussed with cognizant licensee personnel, Observations and Findinas Off-Site Review Committee Activities The team reviewed TS 15.6.5.2 "Off-Site Review Committee," which outlined the OSRC duties and qualification requirements. The OSRC reviewed plant activities and documentation, providing an overview independent of plant management. The team l noted a demonstrated focus on vigorously implementing OSRC goals, in that the OSRC
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had recently recommended two actions to improve the quality of reviews. The OSRC recommended that a representative from OSRC attend key meetings with Institute of
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Nuclear Power Operations and the NRC, and that Committee members be notified of all major plant events to ensure that the OSRC was cognizant of significant issue Nianaaer's Suoervisory Staff Activities i The team reviewed TS 15.6.5.1 " Manager's Supervisory Staff (MSS)," which outlined the l
MSS duties and qualification requirements, as well as the controlling charter for the MSS subcommittee. Due to plant performance improvement efforts, the MSS had been assigned the additional task of generating CRs when draft SEs reviewed by the MSS were determined to be unacceptable. The team reviewed the CRs generated by the MSS, as well as a MSS performance indicator for trending 10 CFR 50.59/72.48 SE The resultant root cause evaluation (RCE) for the rejected SE appeared to be an effective method to preclude recurrence. The expansion of MSS tasks in the area of SEs demonstrated a conservative focus on safety and resulted in heightened engineer awareness of management 10 CFR 50.59 expectations.
l The team attended a MSS Subcommittee meeting held on February 23,1999, and observed that this meeting was conducted in accordance with procedural requirements and in a professional manner. The MSS members raised pertinent questions and l exchanged real time knowledge about issues that were discusse Conclusions The team concluded that the plant review organizations provided the technical expertise l in the multiple areas required to review and evaluate the documents and activities that l occur at the plant. Memoers of the groups were experienced and were aggressive in
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pursuing plant problems and issues. The minutes of the meetings indicated that probing l
questions were asked. The plant review organizations provided an effective and
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vigorous overview of plant activities related to safety.
l- E8 Miscellaneous Engineering issues (92903)
This section describes the review, action and status of selected items which had been identified in previous NRC inspections. Enforcement actions (EA 97-347 and EA 97-505) for the escalated enforcement issues discussed below were addressed in inspection Report No. 50-266/97022(DRP); 50-301/97022(DRP).
E8.1 (Closed) Unresolved item 50-266/301-98013-01: Circuit Breakers Stored in Racked-out l Position. The licensee had not analyzed the mass redistribution effect on seismic qualification with breakers removed from switchgear and had analyzed seismic qualification only for breakers in the racked-in posinon. Subsequent licensee calculations showed that mass redistribution did not negatively affect the switchgear qualification with only a few breakers being racked-out during plant operations. Licensee calculations resolved seismic concems for most 480 VAC and 4160 VAC breakers in the racked-out and test positions; however, in the racked-out position, breaker type DB-75 and the new vacuum breaker were not seismically qualified. Consequently, the licensee added procedural controls for maintaining these breakers in their seismically qualified
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i position and will resolve the seismic issue for the new vacuum breakers. The team concluded that the corrective actions resolved the original concern and no breakers were l observed during the inspection in an unqualified position. This issue is close E8.2 &lgsed) Violation 50-266/301-98010-01: Failure to Evaluate Lost, Damaged, or Out-of-Tolerance Measure and Test Equipment (M&TE). The team reviewed the licensee's corrective actions addressed in letter NPL 98-0323, dated April 30,1998. The actions ensured that installed plant equipment was not affected by the out-of-tolerance M&TE identified in the violation. In addition, procedural guidance for handling M&TE was revised to preclude future recurrence. This issue is close E8.3 (Closed) Violation 50-266/301-98006-07: Failure to Update Battery Loading Calculations. The team verified that calculation N-92-100, " Batteries DOS, D06. D105, D106, D305,10-205 and 2D-205 DC [ direct current) System Master Calculations," and sizing voltage drop and short circuit calculations N-93-056 through N-93-060 were revised based on current design configurations. This issue is close E (Closed) Violations 50-301/98005-01.-02.-03: Inadequate ASME Weld Specifications for: WPS GT-SM/1.1-1PB Revision 4, incorrect Heat input; WPS GT-SM-BU/1.3-1PB, incorrect Heat input; and WPS GT-SM/1.1-1PB Revision 4, Post Weld Heat Treatment Not Specified. The violatione concerned weld procedures used at the plant for the Unit 2 i steam generator replacement that either did not meet ASME Code requirements or did l not prescribe activities performed. The team reviewed the licensee's corrective actions addressed in letter NPL 98-0470 and verified that weld procedures and associated weld ,
procedure qualification tests had been performed, documented, and revised to include j the correct heat inputs as required by the ASME Cod Although the licensee addressed the specific weld procedures stated in the violation, the I response did not appear to fully address actions necessary to prevent recurrence. To resolve this issue, the licensee issued CR 99-807 to address the generic issues of the violation. These generic issues were: (1) Adequacy ofimplementation of ASME Code welding programs; (2) Lack of expertise to ensure adequate technical review of ,
contractor weld procedures; and (3) Adequacy of weld fabrication work monitoring and quality assurance auditing programs to ensure contractor compliance with Appendix B requirement l The team also reviewed two weld procedures used for recent plant modifications, as well I as recent maintenance department welding action items that provided clear ' user friendly' I
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weld procedures for plant welders. No ASME Code violations were identifie Based on the above, the team concluded that sufficient controls were in place to l
adequately address the ASME Code compliance of licensee and contractor procedures used in the plant. These violations are closed.
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E8.5 (Closed) Violation 50-266/301-97023-01: Two Examples of Failure to Follow Procedures. The team reviewed the licensee's corrective actions addressed in letter NPL 98-0214. The first example concerned the issuance of danger tags that were not properly sequenced. The licensee implemer.ted a danger tagging software that prevents printing of a danger tag series without tag sequencing numbers assigned. The second example concerned not performing a prompt operability determination when concerns were identified with relief valve testing as documented on OCR 97-0148. The licensee determined that there were no operability issues with any installed relief valves, and RCE 97-0123 was performed to address the programmatic issues with the CR. This issue is close E8.6 (Closed) Violation 50-301/97023-02: Use of Uncalibrated Stop Watch During Surveillance Testing. Uncalibrated stop watches had been used to measure bus undervoltage relay pickup time delay setpoint values. The team reviewed the licensee's corrective actions addressed in letter NPL 98-0214. As part of the corrective action, the licensee procured additional calibrated stop watches, which were added to the M&TE program to be re-calibrated on a periodic basis. Surveillance procedures were revised to ensure a calibrated stop watch would be used for relay timing. This issue is close E (Closed) Violation 50-266/301-97023-03: Two Examples of Failure to Ensure FSAR Updated. The team reviewed the licensee's corrective actions addressed in letter NPL 98-0214. The first example concerning the replacement of the original plant voltage regulators was incorporated in the FSAR by the June 1998 update. The second example concerned the discharging of radioactive releases from the CVCS holdup tanks versus discharging from the waste disposal system as discussed in the FSAR. The licensee determined that there were concerns with discharging directly from the holdup ;
tanks, such that the process was revised back to that discussed in the FSAR. As such, l no FSAR change was required. To preclude recurrence, the licensee revised NP 10. l to require a prepared and reviewed FSAR revision (if required) prior to approval of an SE by the MSS. This issue is close j E8.8 (Closed) Escalated Enforcement Issue 50-266/301/97022-01: Non-exempt Power Cables Did Not Meet Appendix R Separation Criteria. This item was associated with two existing power cables in the AFW pump room which did not meet the separation criteria in 10 CFR Part 50, Appendix R, Paragraph lli.G.2. The cables had not been included in a previous exemption request for this area. Corrective actions included implementation of fire watches for compensatory measures. In addition, the licensee planed to include the cab:es in an exemption request when the Appendix R program reviews were {
complete. The team verified that fire watch rounds were made for this area. This issue close !
E8.9 (Closed) Escalated Enforcement Issue 50-266/301/97022-03: Electrical Short Circuit
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During a Control Room Fire Could Affect Safe Shutdown. This issue concemed a l condition associated with the potential for postulated fires in the control room to cause
" hot smart electrical shorts" in motor-operated valve (MOV) control circuits. The hot l
short could disable the MOVs and prevent safe shutdown of a reactor. The licensee identified 45 MOVs that could be physically damaged by a hot short, while 14 out of the
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- Potential Common Mode Failure in AFW System Control Circuits. The licensee reported via Licensee Event Report (LER) 50-266/97-040-00 a hypothetical condition associated with the potential for a single failure which could trip two of the three AFW pumps required for Unit 1. The team verified that the licensee had taken adequate corrective actions. This issue is close E8.11 (Closed) Escalated Enforcement issue 50-266/97022-05: Potential Common Mode Failure in Direct Current (DC) Power Supply to AFW. The licensee reported via LER 50-266/97-036-00 a condition associated with the potential for a single failure which could defeat the low pressure suction pressure trip protection logic for three AFW pumps following a seismic or tornado event. The team verified that the licensee had taken adequate corrective actions. This issue is close E8.11 (Closed) Escalated Enforcement issue 50-266/301-97022-06: Non-conservative Setpoint for AFW Pump Low Suction Pressure Trip. The licensee reported via LER 50-266/97-031 a condition associated with the non-conservative setpoint for AFW pump suction pressure trip protective circuits. The proposed corrective actions were to install missile protection for a portion of the AFW suction piping, verify the pump trip time delay relay setpoints, revise the low suction pressure trip setpoint from 6.5 to 7.0 psig, and reduce the acceptance criteria for the turbine-driven pump steam admission valves prior to single unit restart. The team concluded that the corrective actions were sufficient to resolve the original concem and verified that these actions had been taken prior to restart of a single unit. The team also verified that, prior to two unit operation, appropriate plant modifications were implemented to ensure operability of all four AFW pumps. This issue is close E8.12 [ Closed) Escalated Enforcement issue 50-266/301-97022-07: Redundant Safety-related Circuits in the Same Control Board Wireway. This issue was identified by the licensee in LER 50-266/96007-00. The team verified that the licensee had taken adequate corrective actions. This issue is close E8.13 LQlosed) Escalated Enforcement Issue 50-266/301-97022-08: Unanalyzed SW System Alignment. The licensee reported via LER 50-266/97-030-00 a condition associated with the operation of the SW system in a manner which invalidated the system's design basis flow assumptions. Specifically, the SW system had been lined up to more than the two CCW system heat exchangers assumed in the SW flow analysis. This created the potential that necessary SW flow would be diverted from required safety-related loads during postulated accident conditions. The corrective actions included correcting the applicabia system operating procedures and a commitment to operate the SW system in i- accordance with the design basis flow analysis. The team verified that the system l operating procedures (1-SOP-CC-001 and 2-SOP-CC-001) and FSAR Section 9.1, were consistent with the analyzed system alignment. This issue is close l l
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l E8.14 (Closed) Escalated Enforcement Issue 50-266/301-97022-12: Inadequate Surveillance
! of Reactor Trip System Interlocks. The licensee reported via LER 50-266/97-043-00 that l a technical specification change in 1993 had added surveillance requirements and l
intervals for the reactor trip system interlocks, but that these requirements had not been implemented. The team verified that the licensee had taken adequate corrective action This issue is close l E8.15 (Closed) Escalated Enforcement issue 50-266/301-97022-13: Potential Failure of EDG Load Sequencer. The licensee reported via LER 50-266/97-037-00 a condition where the coincidental start of the train 'A' containment spray pump and third train 'A' SW pump could cause actuation of the 480 volt undervoltage protection function. This would result in train 'A' load shedding causing loss of the other train 'A' safety-related pumps about 25 seconds into the loading sequence. The team verified that the licensee had taken appropriate corrective actions. This issue is close E8.16 (Closed) Escalated Enforcement issue 50-266/301-97022-19: Safety injection Delay Times Exceed Design Basis Values. The licensee reported via LER 50-256/97-001-00 an error in the Point Beach accident analysis. Specifically, the assumed initiation times for the high head and low head safety injection did not allow for signal processing, sequencer delay time, or the effects of degraded voltage on the pump starts. The team verified that the licensee had taken adequate corrective actions. This issue is close E8.17 (Closed) Violation 50-266/301-97015-01: Inadequate ELU Surveillance. This violation concerned procedure PBP E-M3-1, " Emergency Lighting," Revision 4, which did not provide adequate guidance to the maintenance staff to perform the quarterly ELU surveillances. The procedure did not specify the use of a calibrated voltmeter, did not include a data sheet that adequately captured the ELU battery and battery charger voltage readings, and did not provide guidance as to what steps were to be taken when battery hydrometer disks indicated potential deficiencies in either the battery or battery charger. The team verified that the procedure had been revised to correct the above mentioned issues. In addition, the team noted that the licensee had taken additional corrective actions to improve ELU performance. The incandescent lights were changed to light emitting diodes to increase the reliability of the indicating lights. The 8-hour discharge test was changed to an ohmic battery test, which provided a qualitative value for trending. Trending information on charger voltage, battery voltage, temperature of battery, electrolyte, and battery impedance was collected and trended in the plant performance trending database. In addition, test jacks were installed in the ELUs to facilitate the use of a voltmeter for testing. This issue is close E8.18 (Closed) Violation 50-266/301-970015-02: Surveillance for ELUs Not Adequately implemented. This violation concerned the failure of electrical maintenance personnel to correctly implement MWP 126 in that the required annual testing of Group IV ELUs was not completed. In addition, electrical maintenance personnel were using the uncalibrated voltmeter located on the front of the ELUs in lieu of a required digital i multimeter. Corrective actions to the violation included installation of test terminals on the front of the ELUs to make testing easier and training staff on the correct method of 27 i l
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l testing ELUs. A review of recent surveillances indicated that they were all completed correctly. This issue is close E8.19 (Closed) Violation 50-266/301-970015-03: Surveillance for ELUs Not Adequately implemented. This violation concerned the failure to correctly implement procedure PBP E-M3-1, which required that the correct alignment of ELUs be verified using existing )
painted markings installed on ELU mounting brackets. Corrective actions to the violation
included ensuring the ELUs were marked and correctly aimed. No mispositioned lights ,
were noted during the inspection. This issue is close j E8.20 (Closed) Violation 50-266/301/97015-04: Fire Brigade Members Not Perforrr!ng Fire Drills. This violation concerned fire brigade members not actively participating in the required two drills per year to remain qualified as fire brigade members. The corrective actions ensured that only those fire brigade members that dress-out in fire fighting equipment and actively participate at the scene of the fire drill were given credit for !
actively participating in the fire drill. In addition, control room operators were removed f from the fire brigade database so these individuals were not considered part of the fire l brigad The team reviewed PC-74 " Conducting and Evaluating Fire Drills," critique forms which indicated that the fire brigade members were no longer given credit for contrcl room coverage as meeting the fire brigade drill requirement. In addition, the team verified that the fire brigade members listed as qualified had participated in the required number of fire brigade drills during 1997 and 1998. The team concluded that the corrective actions appeared effective. This issue is close E8.21 (Closed) Insoection Follow-Up Item 50-266/301-97010-03: The FSAR Stated an incorrect Galvanized Steel Corrosion Rate. Section 5.6.2.1of the FSAR incorrectly listed the galvanized steel corrosion rate. This error resulted in an underestimation in the amount of hydrogen which would be generated from plant galvanized steel following a loss-of-coolant-accident. The team verified that the licensee corrected FSAR section 5.6.2.1 to include the proper galvanized steel corrosion rate. The team also verified that FSAR section 5.6.2.1 no longer stated that the hydrogen produced from galvanized steel l
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corrosion would be insignificant. The inspector reviewed calculation 97-110
" Containment Hydrogen Generation From Documented Sources," and determined that a more conservative hydrogen generation rate was used. The team concluded that the hydrogen generation estimated by the corrected galvanized steel corrosion rate would '
not increase the total post-loss-of-coolant-accident hydrogen generation above the FSAR limits, as confirmed by calculation 97-110. This issue is close E8.22 [Qlosed) Unresolved issue 50-266/301-966010-01: Design Basis Versus Licensing Basis. The NRC had questioned an operability determination that stated control board wiring was a design basis commitment and not a licensing basis commitment, such that separation within the boards was not required. By letter, NPB 97-0100, dated April 16, 1997, the licensee acknowledged that physical separation should be considered a licensing basis requirement for wiring inside the control boards. This issue is close r
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f E8.23 (Closed) Escalated Enforcement issue 50-266/96018-07o: Failure to Correct a Condition Adverse to Quality Regarding Evaluation of Electrical Fault Propagatio Specifically, on June 8,1993, the licensee identified that current-limiting devices on safety-related inverters may not prevent a fault in one circuit from affecting other circuit The team reviewed the response and verified that the licensee had taken adequate corrective actions. This issue is close E8.24 (Closed) Violation 50-266/301-94015-01: Combustible Controls for Hot Work. This violation concerned the failure to follow hot work requirements during plant grinding activities. Combustible materials had been found within 35 feet of grinding activities, and the immediate corrective action was to remove the combustible materials from the hot
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work location. Follow-up actions included improving the training program for hot work and clarifying the guidance in PBNP 3.4.1, " Ignition Control Procedure." In addition, supplemental guidance was given on the identification of combustible materials and possible methods of protecting the job site. The team reviewed 2 years of hot work CRs and determined there had been very few fires as a result of hot work. The improvements in the hot work program were effective. This issue is close i V. Manaaement Meetinas X1 Exit Meeting Summary The team presented the inspection results to members of licensee management at an exit meeting on March 12,1999. The licensee acknowledged the findings presented. The root cause evaluations and self-assessments were identified as proprietary informatio l
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1 PARTIAL LIST OF PERSONS CONTACTED Licensee M. Adrihan, System Engineer L. Armstrong, Manager, Design Engineering J. Brander, Senior Project Specialist, Maintenance Support G. Casadonte, Fire Protection Coordinator D. Evers, System Engineer D. Duenkel, Project Engineer l F. Flentje, Senior Regulatory Compliance Specialist J. Gadzala, Licensing Manager R. Hanneman, Supervising Engineer C. Hill, Operating Supervisor j E. A. Hinshaw, Senior Engineer j N. Hoefert, Manager, Engineering Programs {
R. Hornak, Supervising Engineer, Chemical / Mechanical Design j T. Jessessky, Project Engineer, Electrical System Engineering j V. Kaminskas, Manager, Regulatory Services & Licensing i
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T. Kendall, Mechanical Analysis Group Lead J. Knorr, Regulation & Compliance Manager C. Kramschuster, Design Engineer R. LaRhette, Manager, Nuclear Assurance Section )
i B. McLean, System Engineer J. McNamara, Supervising Engineer, Structural Engineer R. Mende, Plant Manager M. Millen, Supervising Engineer, Nuclear Safety Analysis K. Netzel, Site Design Electrical Lead C. Neuser, IST Engineer K. Oetting, System Engineer M. Omillian, System Engineer R. Pederson, Senior Nuclear Engineer C. Peterson, Director, Engineering L. Peterson, Manager, Engineering Processes j M. Reddemann, Site Vice President 1 M. Rosseau, Supervising Engineer, Modifications J. Schroeder, System Engineer J. Schroeder, System Engineer J. Schweitzer, Site Engineering Manager M. Sellman, Chief Nuclear Officer R. Sier, System Engineer, Electrical Systems K. Skarvan, Electrical Maintenance Planner J. Wilson, Supervising Engineer, Component Engineering M. Woznicki, Engineering Supervisor, Seismic Engineering S. Yuen, Supervising Engineer, Nuclear Systems
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LIST OF INSPECTION PROCEDURES (IP) USED IP 37001: 10 CFR 50.59 Safety Evaluation Program IP 37550: Engineering IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 92903: Followup - Engineering LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Ooened URI 266/301-99005-01 Operability of EDG Room Fire Dampers NCV 266/301-99005-02 Failure to implement Adequate Procedural Guidance for Performance of Independent Verification Closed NCV 266/301-99005-02 Failure to Implement Adequate Procedural Guidance for Performance of Independent Verification URI 266/301-98013-01 Safety-Related Breakers Stored in Racked-out Position VIO 266/301-98006-07 Implementing Adequate VDC Measures /125-VDC System Calculations VIO 301-98005-01 Weld Procedure WPS GT-SM/1.1-1PB, incorrect Heat Input VIO 301-98005-02 Weld Procedure WPS GT-SM-BU/1.3-1PB, incorrect Heat input VIO 301-98005-03 Weld Procedure WPS GT-SM/1.1-1PB, Post Weld Heat Treatment Not Specified VIO 266/301-98010-01 Failure to Evaluate Lost, Damaged, or Out-of-Tolerance VIO 266/301-97023-01 Two Examples of Failure to Follow Procedures VIO 301-97023-02 Use of Uncalibrated Stop Watch During Surveillance VIO 266/301-97023-03 Two Examples of Failure to Ensure FSAR Updated
eel 266/301-97022-01 Non-Exempt PC do Not Meat Appendix R Separation Criteria eel 266/301-97022-03 Electrical Short Circuit During a Control Room Fire Could Affect Safe Shutdown eel 266/301-97022-04 Potential Common Mode Failure in AFW System Control Circuits eel 266/301-97022-05 Potential Common Mode Failure in VDC Power Supply for AFW eel 266/301-97022-06 Non-conservative Set Point for AFW Pump Low Suction Pressure Trip eel 266/301-97022-07 Redundant Safety-Related Circuits in Same Control Board Wireway eel 266/301-97022-08 Unanalyzed Service Water System Alignment
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eel 266/301-97022-12 Inadequate TS Surveillance of Reactor Trip System interlocks eel 266/301-97022-13 Potential Failure of EDG Load Sequence eel 266/301-97022-19 Safety injection Delay Times Exceed Design Basis Values VIO 266/301-97015-01 Inadequate ELU Surveillance
. VIO 266/301-97015-02 ELU Surveillance Not adequately implemented VIO 266/301-97015-03 ELU Surveillance Not adequately implemented VIO 266/301-97015-04 Fire Brigade Members Not Performing Data IFl 301-97010-03 Review Approved Calculation No.97-110 eel 266/301-96018-07o Corrective Action Problems URI 266/301-96010-01 Design Basis Versus Licensing Basis Interpretation VIO 266/301-94015-01 Combustible Controls for Hot Work
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LIST OF ACRONYMS AFW Auxiliary Feedwater ASME American Society of Mechanical Engineers CFR Code of Federal Regulations CR Condition Report CCW Component Cooling Water CVCS Chemical and Volume Control System DBA Design Basis Accident DC Direct Current DRS Division of Reactor Safety eel Escalated Enforcement item EDG Emergency Diesel Generator ELU Emergency Lighting Unit EPRI Electric Power Research Institute FSAR Final Safety Analysis Report IFl Inspection Follow-up item IP inspection Procedure IST Inservice Test LER Licensee Event Report MOV Motor-Operated Valve MR Modification Request l MSS Manager's Supervisory Staff MSSM Manager's Supervisory Staff Subcommittee Meeting Minutes l M&TE Measurement and Test Equipment i
NCV Non-Cited Violation NRC Nuclear Regulatory Commission l
OD Operability Determination l OE Operating Experience l
OSRC Off-site Review Committee QA Quality Assurance QCR Quality Condition Report RCE Root Cause Evaluation SCR Safety Screening Review SE Safety Evaluation SOER Significant Operating Experience Report STPT Setpoint SW Service Water TM Temporary Modification TS Technical Specification URI Unresolved item USO Unreviewed Safety Question VAC Volts Attemating Current VDC Volts Direct Current VIO Violation
- F Degrees Fahrenheit
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l 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 team 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. Inclusion of a document in this list does not ,mply l NRC acceptance of the document, unless specifically stated in the inspection repor Ergcedures l DG-101 Design Guideline, instrument Setpoint Methodology, Revision 2 DG-G02 Guidelines for IWPs, Revision 7,7/11/97 DG-M16 Design Guideline for ASME Section XI Design Reconciliation, Revision 1,1/30/98 NDE 754 Visual Examination (VT-3) of Nuclear Power Plant Components, Revision 8, 11/4/98 NP 1.9. Ignition Control Permit Procedure, Revision 2,8/29/97 NP 2. Independent Verification and Concurrent Checks, Revision 0,7/8/98 NP 5. FSAR Revision, Revision 5,12/9/98 !
NP 5. Condition Reporting System, Revision 11,1/6/99 NP 5. Industry Operating Experience Review Program, Revision 8,12/23/98 l NP 5. Operability Determination, Revision 6,7/29/98 NP 5. Corrective Action Program, Revision 3,10/14/98 NP 5. Open item Tracking Systems, Revision 5,10/14/98 NP 7. Engineering Advisory Committee, Revision 0,1/6/9 ,
NP 7. Modification Requests, Revision 3,5/29/98 !
NP 7. Design Control, Revision 5,9/30/98 )
NP 7. Engineering Change Requests, Revision 3,7/29/98 )
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NP 7. Engineering Change Process, Revision 0,4/8/96 NP 7. Temporary Modifications, Revision 9,4/14/98 NP 7. PBNP Setpoint Changes, Revision 3,7/26/96 ;
NP 7. Contractor Procedures, Revision 1,11/26/97 q l NP 8. Measurement and Test Equipment, Revision 4,12/23/98 j NP 10. Authorization of Changes, Tests, and Experiments (10 CFR 50.59 and 72.48 l Reviews, Revision 11,12/9/98 )'
NP 11. Qualification of Quality Verification Audit / Surveillance Personnel, Revision 3, 1/20/99 NP 11. Quality Verification Audits and Surveillances, Revision 9,2/17/99 NP 11. Internal Assessment Program Coverage, Planning, Scheduling, and Reporting, Revision 5,2/17/99 NP 11. QA Program Significant issues, Revision 2,1/29/99 NP 11. Escalation of Open OCR's, Revision 2,1/29/99 l
NP 1 Work Monitoring Program, Revision 0,2/17/99 :
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OEG.001 Root Cause Evaluation Guide, Revision 2,8/20/98 l SEM System Engineering Handbook, Revision 1,12/10/98 l SEM Site Engineering Group General Expectations, Revision 1,9/11/96 SEM Conduct of Operations - Plant Support Engineering, Revision 0,6/24/94 SEM Conduct of Operations - Plant Reactor Engineering, Revision 4,7/15/98 SOP-CC-001 CCW Operating Procedure, Revision 0,11/6/98
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WP-1 Welding Procedure for Carbon Steels Group P1 to P1 GT-SM, Revision 7,1/8/91 WP-2 Welding Procedure for Austenitic Stainless Steels ASME Group P8 GT-SM, Revision 6,1/8/91 i Form PBF-0036 Wire Removal Log, Revision 3,2/18/97 Surveillance Procedurqs DCS 3. Emergency Diesel Generator Operability, Revision 11,1/11/99 FEP 4.13 Emergency Diesel Generator (G-01/G-02) and Compressor Rooms, Revision 7, 12/10/98 IT 008 IST Stroke Time (Acceptance Criteria), Revision 2,7/3/98 IT 008A IST Stroke Time (Acceptance Criteria), Revision 3,7/3/98 IT 09A Cold Start of Turbine Driven Auxiliary Feed Pump and Valve Test Unit 2 (Quarterly),
Revision 18,8/14/98 MWP 126 Emergency Lignting Performance Monitoring Test, Revision 6 RMP 57 Fire Barrier Penetration Fire Seal Surveillance, Revision 6,4/8/97 PC 75, Part 7 Semiannual G-02 Diesel Generator Fire Damper and Ventilation Surveillance Test, Revision 1,1/30/97 l
Audits and Self-Assessments J A-P-97-13 Inservice inspection Program (Site Engineering),5/19-6/11 and 7/7-18/97 !
A-P-97-14 Maintenance Rule Audit, S/12-22/97
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A-P-97-24 Fire Protection Program,8/18-27/97 l A-P-98-01 Design Engineering,4/20-5/21/98 l
A-P-98-02 Inservice inspection and Repair / Replacement Programs,3/16-4/16/98 !
A-P-90-17 Corrective Action and Operating Experience Programs 1/11-2/5/99 S-P-97-13 10CFR50.59 Screenings and Evaluations, 12/8-16/97 S-P-97-16 G-03 and G-04 Emergency Diesel Generator Close-out Surveillance,9/16/97 SA -98-16 Point Beach System Engineering Self-Assessment for July 20-24,1998,8/17/98 WMR 99-0023 Quality Verification Mini-assessment of Temporary Modifications Modifications 94-091*E Unit 1 Reconfiguration of Supply and Return CCW Flex Hosing 96-071 *D install Cables and Raceways for GO2 Governor 97-099*A-E Unit 2 AFW Trip / Throttle Valve Upgrade j 97-085*A, B Unit 2 Add Motor Operators to SI-857A(B) ,
97-129*B Unit 2 Replace Check Valves AF-100 & AF-101 l 98-048 Unit 2 Pressurizer Level Controller Upgrade 98-116 Unit 2 4kV Breaker Change-out to Vacuum-Style Breakers 93-025*B, E, F Unit 1 Main Control Board Modifications i 95-041 *B Unit 1 Ourpressure Protection of 1RH700/701/720 95-048 Unit 1 EDG Output Breaker Undervoltage Permissive Time Delay l l
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96-064*B Unit 1 Upgrade SW Piping Supports Outside Containment j 97-068*A Unit 1 Charging Pump Vari-drive i 97-135 Unit 2 Drill 1/8" in Upstream Disc of 2CV-129 Valve ;
90-015*A GO2 Sump Tank Level Switch 35 I l
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o-l 97-077 Replacement of DG-02 Lube Oil Piping with Flex Hose !
97-112*A Unit 0 G-01/G-02 Fire Damper Modification S_gtooint Chances i
STPT Reactor Trip OPAT, Unit 2, Revision 12 l STPT Feedwater Flow and Steam System Alarms, Revision 4 STPT 14.11 Setpoint Change AFW, Revision 15 STPT 2 Safety-Related Inverters and Instrument Busses, Revision 0 STPT 2 Safety-Related Battery Chargers and DC Busses, Revision 0 Temocrary Modifications
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TM 96-023 Repair of Valve SW-660 TM 98-020 Replace Resistance Temperature Detector for TR-2001 on A Reactor Coolant Pump TM 98-049 Temporary Air Filters for Air Compressor K-2A TM 98-034 Swap Govemor Valve #3 & #4 Control Card TM 98-036 Install Kill Switch for 2P-37C & D l TM 98-041 Disconnect Transformer 2X02 TM 99-006 Temporary Heat Trace for Refueling Water Storage Tank Line TM 98-014 Install High Capacity Sump Pump TM 98-045 Valve FD-00050 Furmanite Repair Operatina Exoerience Documentation OE 9587 4160 Volt Breaker Anomaly OE 8942 Alarms Disabled By improper Recorder Operation OE 8714 Inadvertent Dilution During Low Power Operation at Vogtle OE 8924 Letdown Filter Replacement Results in Personnel Contamination j OP 9176 Operator Proficiency Watches Activities not in Compliance with 10CFR50.54/55 OE 8803 Temporary Power Cord Violates Separation Criteria OE 9215 High Radiatior: Levels During Refueling Cavity Floodup SOER 93-02 Circuit Breaker Reliability PS 33289 Steam Generator Blowdown and Sample Isolation Valves Not Tested Per Generic Letter 96-01 MR-H-98-036 Part 21 - Failed Relay Because of Insufficient Solder Root Cause Evaluations RCE 97-0123 Engineering Programs Failed,4/3/98 RCE 97-134 CCW Flow Instrument inaccurate, Revision 0,2/98 RCE 98-0014 Unexpected Service Water Isolation During PBTP-77,4/21/98 RCE 98-0025 Training Nuclear information Management Backlog,5/1/98 RCE 98-0078 1SI-00896A Rework Due to incorrect Reassembly and Testing Sequence,8/12/98 RCE 98-0148 P-38A AFW Pump Recirculation Valve Found Failed Shut,2/3/99 RCE 98-0150 Unit 1 Turbine Driven Auxiliary Feedwater Pump Turbine Maintenance RCE 98-0151 Missed TS Surveillance on Unit 1, Safety injection Valves SI-852A/B,9/4/98 RCE 98-1360 Unit i Turbine-Driven AFW Pump Turbine Maintenance Rework,10/30/98 l 36 .
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10 CFR 50.59 Safety Evaluations (SEs)
, SE 96-077-02 Relocation of G-01/G-02 Fuel Oil Level Switches, 12/11/97 l SE 97Jt5 1(2)RH-700/701/720 Banner Pressure Relief and Overpressure Protection,5/16/97 SE 97-090 Screening and Safety Evaluation for Modification 95-041*B, Capping of Leak-Off Lines of 1/2RH-700,1/2RH-701 and 1/2RH-720,6/6/97 SE 97-191-01 Service Water Pipe Support Modifications (Unit 1 Outside Containment)- Revised Thermal Mode and Hydraulic Loads,2/5/98 SE 97-207 Screening and Safety Evaluation for Modification 97-099*E,12/16/97 SE 97 208 IT-08A/09A, Cold Start Testing of Turbine-Driven Auxiliary Feedwater Pump and Valve Test Unit % (Quarterly), 12/16/97 SE 98-003-01 Service Water Return Header Hot Tap for Repair of 1SW-2907/2908 SE 98-07 Main Control Board 1c04 Maintenance to Assure Proper Control Wire Separation for the Unit 1 Chemical and Volume Control System,2/13/98 SE 98-011-01 Screening and Safety Evaluation for Modification 97-112*A SE 98-013 Generic Letter 96-01 Pushbutton Modification,2/3/98 SE 98-019 Main Control Board C01 Maintenance to Assure Proper Control Wire Separation for the Service Water System. 2/13/98 SE 98-020 Main Control Board 1C03 and C01 Maintenance and Modification to Assure Proper Control Wire Separation for the Main Steam System,2/13/98 SE 98-021 Unit 1 and Unit 2 Common Main Control Boards C01 and CO2 Maintenance to Assure Proper Control Wire Separation for the 480 VAC System,2/13/98 SE 98-022 Unit 1 Main Control Board 1C03 Maintenance to Assure Proper Control Wire Separation for the Component Cooling System,2/13/98 SE 98-023 Unit 1 and Common Main Control Boards C01,1C03 and ic04 Maintenance to Assure Proper Control Wire Separation for the Reactor Coolant System,2/13/98 SE 98-024 Unit 1 and Unit 2 Common Main Control Board C02 Maintenance to Assure Proper Control Wire Separation for the 4160 VAC System,2/13/98 SE 98-027 Unit 1 and Common Main Control Boards C01, CO21C03 and 1c04 Maintenance to Assure Proper Control Wire Separation for the Unit 1 125 VDC System,2/24/98 SE 98-028 Main Control Board 1C03 and C01 Maintenance and Modification to Assure Proper i Control Wire Separation for the Auxiliary Feedwater System,2/24/98 I SE 98-029-01 Main Control Board C01 Maintenance to Assure Proper Control Wire Separation for the Safety injection System,3/5/98 SE 98-030 Unit 1 Main Control Board 1C03 Maintenance to Assure Proper Control Wire Separation for the Residual Heat Removal System,2/24/98 SE 98-034 Unit 1 Main Control Boards 1C03 and 1C04 Maintenance to Assure Proper Control Wire Separation for the Unit 1 120 VAC System,3/5/98 SE 98-035 Unit 1 and Common Main Control Boards C01,1C03 and 1C04 to Assure Proper Control Wire Separation for the Unit 1 Engineered Safety Features System,3/5/98 SE 98-036 Unit 1 Main Control Board 1C03 Maintenance and Modification to Assure Proper Contro! Wire Separation for the Unit 1 Condensate & Feedwater System,3/5/98 SE 98-039 Unit 1 Steam Generator Seal Plate Modification,3/19/98 SE 98-042 Unit 1 Main Control Board 1C03 Maintenance and Modification to Assure Proper Instrumentation Wire Separation for the Unit 1 Feedwater System. 3/20/98 SE 98-044 Unit 1 and Common Main Control Boards C01,1C03, and 1C04 to Assure Proper Control Wire Separation for the Unit 1 Reactor Protection System,3/20/98 SE 98-045 CCW lsolation in Emergency Operating Procedure's,3/20/98
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l SE 98-049 Unit 1 Main Control Board 1C04 Maintenance to Assure Proper Control Wire
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Separation for the Unit 1 Nuclear instrumentation System,3/25/98 SE 98-050 Unit 1 and Common Main Control Board C01 Maintenance to Assure Proper Control Wire Separation for the Unit 1 Containment Cooling System,3/25/98 SE 98-051 Unit 1 Common Main Control Board C01 Maintenance and Modification to Assure Proper Control Wire Separation for the G01/G02 Fuel Oil System,3/25/98 SE 98-055 Unit 1 Main Control Board ic04 Mainteriance to Assure Proper Control Wire Separation for the Unit 1 Containment Purge Supply and Exhaust System,3/31/98 SE 98-057-02 Diesel Generator Breaker Closure Delay Modification - Unit 1,6/26/98 SE 98-072 Screening and Safety Evaluation for Modification 97-112*A SE 98-093 Function Change for Service Water Isolation Valve,6/17/98 SE 98-097 Revisions to Service Water Pump Inservice Test Flow Criteria,7/8/98 SE 98-111-01 Screening and Safety Evaluation for Modification 98-116 SE 98-112-02 Reactor Vessel: Evaluation of New Installed Baffle Former Pattern & Design SE 98-116 Resolving the Pressure Locking of CVCS Gate Valve 2CV-1299,9/10/98 SE 98-171 Convert 2SI-857A(B) to Motor-Operated Valves and 2SI-897A(B) to Fail Open, 11/30/98 SE 98-176 Removal of SI-885B,11/25/98 SE 97-207 Auxiliary Feedwater Valve and Instrument Loop Modification l
Safety Screenina Reviews (SCRs) !
SCR 97-122C Obtain "As-Built"information for safety related control wires of Main Control Board Panels 1C03 and 1C04,9/30/97 SCR 97-694 Provide SILTEMP Wrapping, for Separation, of Train a Wires of the Steam Generator Blowdown isolation Valve I/K5958 Control Switch Located in the Main Control Board 1C03,6/9/97 SCR 97-926 Screening and Safety Evaluation for Modification 97-077 SCR 97-1752 Screening for STPT 14,11,11/15/97 SCR 97-1871 Screening and Safety Evaluation for Modification 97-009'E l SCR 97-2749 Screening and Safety Evaluation for Modification 96-071*D,10/1/97 SCR 98-0001 Screening and Safety Evaluation for Modification 97-009'E SCR 98-0034 IT 03A/04A,1/20/98 SCR 98-0072 Ol 130, Performance Test of 1HX-15D1-D8 Containment Fan Cooler,1/20/98 SCR 98-0088 Procedure Change to OP 11 A, EDG G-01 (G-02),1/22/98 SCR 98-0089 Procedure Change to OP 118, EDG G-03 (G-04),1/22/98 SCR 98-0132 Temporary Ctange to IT 07A for Post Modification Test of P-32A Service Water Pump,1/30/98 SCR 98-0213 Delete the Two Unused Wires Internal to the Main Control Board C02 for Components Associated with MOB-222 listed in WO No. 9712752 to Comply with the Requirements of the Design Guide DG-E07 for Proper Train Separation,2/11/98 SCR 98-0658 Temporary Change to IT 07A for Post Modification Test of P-32A Service Water Pump,4/29/98 l SCR 98-0719 Modification 97-068*A, Charging Pump Controller Nitrogen Backup Fitting
' SCR 9'8-0987 Screening and Safety Evaluation for Modification 93-025*F SCR 98-1021 Screening and Safety Evaluation for Modification 94-091*E,6/9/98 SCR 98-1122 NP1.2.5, Special Test Procedures, Revision 5,6/18/98 SCR 98-1294 NP 5.3.7, Operability Determinations, Revision 6,7/23/98
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e SCR 98-1402 Screening and Safety Evaluation for Modification 91-130,8/17/98 SCR 98-1625 Screening and Safety Evaluation for Modification 98-048,8/28/98 l SCR 98-2080 Screening and Safety Evaluation for Modification 97-129'B,11/2/98 Operability Determinations CR 96-0650 Reserve Water Storage Tank Recirculation Pipe Not Seismic CR 97-1361- EDG Calculation Has Unclear Assumptions CR 97-1822 Brass Valves in Waste Gas System, Revision 0,3/12/99 CR 97-1838 Appendix R Scenario Disables CCW and Charging Pumps l
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CR 97-2036 Appendix R Safe Shutdown Capacity in AFW Pump Room CR 97-3913 CCW Flow Instrument Accuracy CR 98-0564 CCW Susceptibility to Line Break Inside Containment, Revision 4,6/15/98 CR 98-1629 Isolated SW, SI, CCW, CVCS, Reactor Coolant Piping Sections in Containment
- CR 98-1636 2SI-825-B Motor-Operated Valve, Revision 0 .
CR 98-2172A CCW Flows to Heat Exchanger !
CR 98-2253 CVCS Valves Not inservice Test in Closed Position CR 98-2556 SW Fouling Factors .
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CR 98-2572 Safety injection Piping and Supports Could Degrade at High Sump Temperatures '
l CR 98-2704 Calculation to Raise SW Temperature to 78 *F l LCR 98-3399 Calculation for EDG Heat Exchanger with Elevated SW Temperature i CR 98-3754 Main Steam HELB at Steam Generator Nozzle May Not Meet Criteria for Whip Restraints CR 99-0063 AFW Pump Shaft CR 99-0604 48.5 *F SW Limit on U-1, Revision 2,2/22/99 QMgulations 90-0015 Response Time for 3/8 inch Break in Reactor Coolant System, Revision 0 97-0110 Containment H2 Concentration Calculation,7/4/97 97-0147 Water Volume in Seismic Piping Upstream of AFW Pumps, Revision 2,7/11/97 97-015 AFW Supply Missile Protection 97-0172 Available Water to AFW Pumps with Pipe Break at Elevation 25-6 97-0215 Water Volume Swept by all Four AFW Pumps Following a Seismic / Tornado Event Affecting Both Units, Revision 1,1/11/98 97-0231 AFW Pump Low Suction Pressure Loop Uncertainty, Revision 0,12/1/97 97-0235 G-02 Standpipe Isolation Valve 97-025 Overload Heater Sizing For Motor Protection of AFW Motor-Operated Valves 1MS-
' 2082 and 2MS-2082,12/29/97 96-0285 Flow Uncertainty Associated w/WMTP-12.11, Revision 0,12/18/97 98-0076 Modification 94-091*E Flex Hose Installation,6/11/98 98-0119 Seismic Evaluation of 4160 Volt Switchgear, Revision 1 99-0011 Seismic Evaluation of 1 A-05, Cubicle 1 A52-66 and 2A-05, Cubicle 2A52-67 for New Vacuum Breaker Installation, Revision 0 N-92-100 Batteries DOS, D06, D105, D106 and 2D-205 DC System Master Calculation, Revision 2,11/23/98 i N-94-130 4160 Volt and 480 Volt Safeguards Buses Loss of Voltage Relay,5/31/95 l N-94-158 AFW Design Flows, Revision 2
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, PBNP-1 C-25 Steam Generator Narrow Range Level, Revision 4,6/12/98 WE 200103 Piping System Qualification Report, Revision 0,12/11/97
- WE-100164 Instrument Air Tubing Downstream of Valve 1 A-112, Revision 0,5/14/98 WE-300035 Piping System Qualification Report, Revision 1,9/30/97 M09334-251-RH1 Residual Heat Removal Line Pressurization, Revision 1,4/13/98 M09334-425-AF1 Evaluation: Impact of Check Valve AF-100/101 Replacement, Revision 0 E-9334-388-DG1 Medium / Low Voltage Decay Profile for a Loss of Source Condition, Revision 1
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Manaaer's Supervisory Staff Subcommittee Meetina Minutes Manager's Supervisory Staff Subcommittee Meeting Minutes: MSSM 98-001, MSSM 98-002, MSSM 98-003, MSSM 98-004, MSSM 98-007, MSSM 98-008, MSSM 98-011, MSSM 98-012, MSSM 98-014, MSSM 98-017, MSSM 98-019, MSSM 98-020, MSSM 98-025, MSSM 98-032, MSSM 98-038, MSSM 98-039, MSSM 98-041, MSSM 98-052, MSSM 98-065, MSSM 98-069, MSSM 98-079, MSSM 98-097, MSSM 98-105, MSSM 98-113, MSSM 98-119 MSSM 98-124, MSSM 98-130, MSSM 98-139 Off-site Review Committee Minutes ;
Offsite Review Committee Minutes Nos. 61,62,63,64 10 CFR 50.59/50.72.480SRC Subcommittee Meeting minutes,9/9/98 l
Miscellaneous Documents
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Charter; 10 CFR 50.59 Review Subcommittee of the Point Beach Off-Site Review Committee,12/97 Charter; Point Beach Nuclear Plant Off-Site Review Committee,7/22/97 Point Beach Safety Evaluation Steering Committee Minutes,2/5/99 Point Beach NPBU 50.59/72.48 Feedback Newsletter,2/27/98 EPRI TR-106826 " Battery Performance Monitoring by Internal Ohmic Measurements" Memorandum, " Temperature Correction of Battery Ohmic Measurements," from Dan Funk to Wade Berger,3/23/99 Condition Reports
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CR 95-0284 Discrepancies Between Combustible Loading Classifications and Appendix R Exemption Request ;
CR 96-0915 R/R/M and Pressure Test Procedure Providing inadequate Guidance CR 96-1264 Potential Negative Impacts on Several Post-TMl Radiological Evaluations CR 97-0154 Inservice Test Program Missing Some Valves with Safety Functions CR 97-0323 Air-Operated Valve Testing Program Discrepancies CR 97-1632 Instrument Air Upgraded to A(1) per Maintenance Rule CR 97-1822 Improper Brass Fittings in Waste Gas System Do Not meet Code Requirements CR 98-0317 2P-011 A Missing Outboard Bearing Finger Ring CR 98-0654 M&TE Not Functional CR 98-0897 Operating Permit Used for Major Activity CR 98-1475 Motor-Operated Valve Pinion Gear and Pinion Key Damaged CR 98-1636 M&TE Out-of-Tolerance CR 98-1655 Residual Heat Removal Pump Discharge Check Valves Closed Safety Function
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,e CR 98-1913 : Independent Review of WE Design installation Documents for U1R24 CR 98-2119 Planned Maintenance of G05 initiated While G01 is Out-of-Service CR 98-2382 Containment Hydrogen Monitors May Not Be Maintained per Environmental Qualification Requirements-Unit 1 Restart CR 98-2525 Discrepancies Between the Greenline Diagrams and the Q-list in CHAMPS CR 98-2610 incorrect Oil in U-1 AFW Pump Turbine Outboard Bearing (1P-029-T)
CR 98-2634 Missed Surveillance CR 98-3841 ASME Section XI Pressure Tests Not Performed CR 98-3291 1P-28A Steam Generator Feed Pump Suction Socket Welded Elbow Steam Leak Upsteam of 1CS-56 )
CR 98-3341 Bag of Lead Shot Caught on Fire During Cask Structural Lid Welding CR 98-3526 Auto Condensate Pump Start and Heater Drain Tank Dump Valve Cycled Open CR 99-0750 Wire Removal Log is Confusing CR 99-0766 Insufficient Guidance for Disposition of Gaps that Exceed Criteria CR 99-0767 Improper Materials Used on System Described in the FSAR not Fully Evaluated CR 99-0772 Emergency Lighting Procedure Problems ;
SW Inlet Temperatures, Effect on CCW System, 3/10/99
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CR 99-0777 CR 99-0795 Air Entrainment AFW Pumps,3/11/99 CR 99-0797 Two AFW Supports Gaps Exceed Criteria,3/11/99 Quality Condition Reports QCR 96-0040 - NPBU Specification Process inadequacies QCR 98-0196 Discrepancies in ESP Personnel Qualification Documentation QCR 98-0311 RMP 59, Unit 2 Containment Fire Seal Surveillance Performance Requirement l
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