IR 05000266/1997022
| ML20203B145 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/18/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20203B135 | List: |
| References | |
| 50-266-97-22, 50-301-97-22, EA-97-347, EA-97-505, NUDOCS 9802240216 | |
| Download: ML20203B145 (16) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION lli Docket Nos: 50-266,50-301 License Nos: DPR 24, DPR 27 EA Nos: 97 347 and 97-505 Report No: 50-266/97022(DRP); 50-301/97022(DRP)
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- Licensee
- Wisconsin Electric Power Company, WEPCO
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Facility: Point Beach Nuclear Plant, Units 1 & 2 Location: 6612 Nuclear Road Toa Rivers, WI 542419516 Dates: November 4,1997, through February 3,1998 Inspectors: F. Brown, Senior Resident inspector P. Louden, Resident inspector l
Approved by: J. W. McCormick-Barger, Chief, Reactor Projects Branch 7 9002240216 900210 PDR ADOCK 05000266 G PDR
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EXECUTIVE SUMMARY Point Beach Nuclear Plant, Units 1 & 2 NRC inspection Report No. 50-266/97022(DRP); 50-301/97022(DRP)
This specialinspection consisted of a review of 20 licensee event reports (LERs) submitted between September 9,1996, and December 15,1997. The specialinspection was conducted by the resident inspector Enoineerina The inspectors identified 20 LERs which documented desl9n basis conformance problems at Point Beach. Each condition was classified as an example of an apparent violation. The inspectors determined that the licensee had identified each of these conditions and had specified corrective actions which appeared appropriate to the identified condition (Section EB.1)
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Report Details Ill. Ennineerina E8 Miscellaneous Engineering issues E8.1 Review of Licensee Event Reports fLERs) Inspection Scope finspection Procedure flP192700)
The inspectors reviewed 20 LERs submitted between September 4,1996, and Decemt'er 15,1997. This represented approximately one half the LERs submitted during that time frame. These LERs were selected because each dealt with inadequate implementation of the facility design basis, each was the result of performance predating NRC Enforcement Actions (EAs) 96 273 and 97-075, and each dealt with an issue for which the licensee had implemented significant corrective actions at a programmatic level. The inspectors reviewed each LER using the guidance of IP 92700, Observations and Findinas The 20 LERs that the inspectors reviewed were found to fit into 4 categories: design issues associated with the failure to satisfy 10 CFR Part 50, Appendix R, requirements; design issues associated with the failure of the auxiliary feedwater (AFW) system to satisfy its design basis; issues associt.ted with the failure to operate safety-related systems in a manner consistent with the system design basis; and safety-related "
structures, systems, and components (SSCs) affected by the failure to adequately implement or maintain the SSC design basis. Completion of corrective actions for each LER closed below will be tracked against an Escalated Enforcement item (EEI) numbe For each LER ravit.wed, the inspectors found that the LER accurately described the non-conforming c9nditio (1) Appendix R issues: three LERs dealt with aspects of the facility which did not meet the safe shutdown requirements of Appendix R. The inspectors considered the conditions described in these LERs to be closely associated with the apparent violations of 10 CFR Part 50, Appendix R, documented as Eels 97010-04, 05,
-06, and -07 in Inspection Report No. 50-266/97010(DRS); 50-301/97010(DRS).
fClosed) LER 50-266/97 033 00: Non-exempt Power Cables do not Meet Appendix R Separation Criteria. This LER described a licensee-identified condition associated with two existing power cables in the AFW pump room which did not meet the separation criteria in 10 CFR Part 50, Appendh Paragraph lil.G.2. This condition was identified as the result of a licensee self-assessment of the Appendix R program. Corrective actions included implementation of compensatory measures and submittal of an exemption request for the two cables. The failure to identify and address the lack of cable separation for these two cables in the AFW pump room during the original Appendix R reviews was an example of an apparent violation (eel 50 266/97022-01(DRP);
50-301/97022-01(DRP)) of 10 CFR Part 50, Appendix R, Paragraph lil. ,
(Closed) LER 50 266/97-032-00: Inadequately Rated Electrical Buses Could Disable Switchgear and Cause Secondary Fires. This LER described a licensee-identified cendition associated with the potential for bolted three-phase faults to exceed the capability of required switchgear, including cucent interrupting
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devices. This condition was identified as the result of a licensee self assessment of the Appendix R program. Corrective actions included implementation of compensatory measures, completion of additional technical evaluations, and performance of modifications or design changes. The failure to identify and address the potential for bolted three-phase faults during the original Appendix R reviews was an example of an apparent violation (eel 50-266/97022-02(DRP);
50-301/97022-02(DRP)) of 10 CFR Part 50, Appendix R, Paragraph Ill. (Closed) LER 50-266/97-022-00: Electrical Short Circuits During a Control Room Fire Could Affect Safe Shutdown Capability. This LER described a licensee-identified condition associated with the potential for postulated fires in the control room to cause " hot smart electrical shorts" in motor-operated valve control circuits that could disable tne valves et prevent safe shutdown of a Unit. This condition wat identified as the result of u dicensee self assessment of the Appendix R program. Corrective actions consisted of modification of all affected motor-
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operated valve control circuits to eliminate the potential for bypassing the limit l
switches. The failure to identify and address the potential for hot smart electrical shorts during the original Appendix R reviews was an example of an apparent violation (eel 50-266/97022-03(DRP); 50-301/97022-03(DRP)) of 10 CFR Part 50, Appendix R, Paragraph lli. (Closed) Unresolved item (URI) 50-266/96018-21(DRS): 50-301/96018 21(DRS!:
Hot Smart Short Potential. This URI dealt with the hot smart short concem discussed in LER 50-266/97-022 00. The associated inspection report documented that the licensee's original response to the issue was not sufficiently aggressive or tedWcally rigorous, but that appropriate reviews were underwa The content of the above LER demonstrated that the subsequent assessments were sufficiently rigorous and conservativ (2) AFW System issues: four LERs dealt with problems associated with the design of AFW pump protection circuit (Closed) LER 50-266/97 04100: Potential Common Mode Failure in AFW System Control Circuits. This LER described a licensee-identified condition associated with the potential of a single failure which could trip two of the three AFW pumps required for Unit 1. The unaffected pump could not supply the required flow to a Unit under all accident conditions. This potential common mode ,
failure was the result of running protective circuit cables for the turbine-driven AFW pump and one motor driven AFW pump in a common conduit. The cables were run in response to NUREG 0737, " Clarification of TMI (Three Mile Island)
Action Plan Requirements." This condition was identified as the result of a licensee self assessment of the AFW system. Corrective actions consisted of modifications to the control circuitry for the two AFW pumps. The failure to provide single failure protection, that is, cable separation, for the AFW pump protective circuitry installed in response to NUREG-0737 was an example of an apparent violation (eel 50-266/97022-04(DRP); 50-301/97022-04(DRP)) of 10 CFR Part 50, Appendix B, Criterion Ill, " Design Control," which requires, in part, that design changes be controlled to assure that the design basis is correctly translated into specifications, drawings, procedures, and instructions.
' (Closed) LER 50-266/97-036-00: Potential Common Mode Failure in DC [ Direct Current) Power Supply for AFW, This LER described a licensee-identified condition associated with the potential of a single failure v/hich could defeat the
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low suction pressure trip protective logic for three AFW pumps following a seismic or tomado event that caused the loss of the normal APN suction source. The unaffected pump could not supply the required AFW flow to both Units. This potential common mode failure was the result of supplying power for the low pressure suction trip for three AFW pumps from a common power source. The low pressure suction trip was installed in response to NUREG-0737. This condition was identified as the result of a Nensee self assessment of the AFW system. Corrective actions consisted of modifications to the low suction pressure trip circuitry for the AFW pumps. The AFW system was considered operable for single Unit operation only until the required modifications are completed. The failure to provide single failure protection, that is, power source separation, for the AFW pump protective circuitry installed in response to NUREG-0737 was an example of an apparent violation (eel 50-266/97022-05(DRP);
50-301/97022-05(DRP)) of 10 CFR Part 50, Appendix B, Criterion Ill, which requires, in part, that design changes be controlled to assure that the design basis is correctly translated into specifications, drawings, procedures, and instruction [ Closed) LER 50-266/97-03100: Nonconservative Setpoint for AFW Pump Low Suction Pressure Trip. This LER described a licensee-identified condition associated with the nonconservative setpoint for AFW suction pressure trip protective circuits. This condition could have led to cavitation in all AFW pumps and resulted in the pumps not being able to meet their design functions. This potential common mode failure was the result of falling to include instrument uncertainty in the original setpoint calculations and the failure to consider seismically induced loss of nonsafety related pump suction piping. The protective logic was installed in response to NUREG-0737. This condition was identified as the result of a licensee self assessment of the AFW system. Corrective actions consisted of modifications to the system and additional testing of system component:1 to bound acceptsble pump trip delays. At the completion of this special inspection, new turbine throttle trip valves were being installed to provide two-Unit resolution of this issue. The failure to provide an appropriate setpoint for the AFW pump protective circuitry installed in response to NUREG-0737 was an example of an apparent violation (eel 50-266/97022-06(DRP);
50-301/97022-00(DRP)) of 10 CFR Part 50, Appendix 0, Criterion lil, which requires, in part, that design changes be controlled to assure that the design basis is correctly translated into specifications, drawings, procedures, and instruction (Closed) LER 50 266/96-007-00: Redundant Safety-related Circuits in the Same Control Board Wireway. This LER described a licensee-identified condition associated with redundant AFW pump control circuits routed in the main control boards without adequate separation and without adequate fault current interruption protection. The licensee determined that a single fault in one train of AFW could cause circuit damage to the other AFW train equipment. All three AFW pumps associated with Unit 1 were declared inoperable until certain power supplies for one train could be de-energized. This condition was identified as the result of a licensee self assessment of the AFW system. Corrective actions consisted of modifications to the fault current interruption protection and permar.ent modifications to establish control circuit separation. The operation of Unit 1 from the time of initial criticality through identification of the single fault induced common mode failure of redundant trains of AFW and declaration of pump inoperability on August 14,1996, was an example of an apparent violadon
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(eel 50 266/97022-07(DRP); 50-301/97022 07(DRP)) of Technical Specification (TS) 15.3.4.A.2, which requires, in part, that both trains (including three pumps) of AFW be operable prior to making a Unit critica (3) Operation of Systems: four LERs dealt with the failure to operate safety-related systems in a manner consistent with the design basi (Closed) LER 50-266/97-030-00: Unanalyzed Service Water (SW) System Alignment. This LER described a licensee identified condition associated with the operation of the SW system in a manner which invalidated the design basis flow -
assumptions for that system. Specifically, the SW system had been lined up to l
more than the two component cooling water (CCW) system heat exchangers assumed in SW flow analysis.- This created the potential that necessary SW flow would be diverted from required safety-related loarts during postulated accident conditions. This condition was identified as the result of a licensee self-l assessment of the LW system. Corrective actions consisted of modifications to l the applicable system operating procedures and a commitment to operate the SW
system in accordance with design basis flow analysis. The operation of the SW system in a manner inappropriate to the circumstances, that is, with flow distribution that was inconsistent with the design basis, was an example of an apparent violation (eel 50-266/97022 08(DRP); 50-301/97022-08(DRP)) of 10 CFR Part 50, Appendix B, Criterion V," Instructions, Procedures, and Drawings," which requires, in part, that activities affecting quality be implemented in accordance with procedures which are appropriate to the circumstance (Closed) LER 50-266/97-026-00: TS Violation of Operability Requirement for Main Steam Line Isolation. This LER described a licensee-identified condition associated with the operation of the reactor coolant system (RCS) temperature instrumentation in a manner that was inconsistent with the design basis and rendered the main steam line isolation function of the reactor protection system (RPS) inoperable. Specifically, the licensee installed scaling resistors in the RCS temperature instrumentation to maintain indication on scale. These scaling resistors rendered the protective function of main stream line isolation inoperable unless the temperature input was in " trip " This condition was identified as the result of a licensee self-assessment of operating procedures. Corrective actions consisted of revising applicable procedures to direct that the RCS temperature input to the RPS main steam line isolation function be placsd in the trip condition prior to leaving cold shutdown when scaling resistors are installed. The operation of each Unit above cold shutdown conditions with scahng resistors installed in the RCS temperature instrumentation and the temperature input to the RPS main steam line isolation logic not in trip was an example of an apparent violation (eel 50-266/97022-09(DRP); 50-301/97022-09(DRP)) of TS Table 15.3.5-4, item 2.b, which requires that the affected Unit be placed in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if this RPS function is inoperabl (Closed) LER 50-266/97 025-00: Pressurizer Level Controlled Higher than Assumed in Accident Analysis. This LER described a licensee-identified condition associated with the operation of the RCS pressurizer at a level greater than the automatic zero load setpoint used in accident analysis. Specifically, the pressurizer was operated at 30 percent indicated level by placing the level controller in manual. The assumed start point for low power accident analysis was 20 percent indicated level. Three operating procedures had been changed to allow this manual control of pressurizer level. This condition was identified as the
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result of a licensee self assessment of operating procedures at Point Beac Corrective actions consisted of modifications to the applicable system operating procedures and a commitment to perform a root cause evaluation of the inappropriate procedure changes associated with this condition. The operation of the RCS in a manner inappropriate to the circumstances, that is, with pressurizer level that was inconsistent with the design basis, was an example of an apparent violation (eel 50-266/9702210(DRP); 50-301/9702210(DRP)) of 10 CFR Part 50, Appendix B, Criterion V, whicn requires, in part, that activities affecting quality be implemented in accordance with procedures which are appropriate to the circumstance (Closed) LER 50-266/97-009-00: Potential Safety inje : tion (SI) Failure During Filling of St Accumulator. This LER described a licensee identified condition associated with a procedure that allowed the operation of either high head Si pump to fill the Si accumulators while a unit was at power. A licensee review of this practice determined that the use of the "A" train SI pump to fill the accumulators rendered the "A" train pump inoperable and that the use of the
"B" train Si pump rendered both trains of Sl inoperable. The SI pumps were considered inoperable because the potential to divert SI flow from the core into the accumulators had not been evaluated in the systems design basis. This condition was identified as the result of a licensee self assessment of operating procedures. Corrective actions consisted of revising applicable procedures to direct that the "B" train Si pump not be used, and that the "A" train SI pump be declared inoperable when it was used to fill the SI accumulators for an operating unit. The operation of the SI system in a manner inappropriate to the circumstances, that is, creating potential flow diversion paths not analyzed in the design basis, was an example of an apparent violation (eel 50-266/97022-11(DRP); 50 301/97022-11(DRP)) of 10 CFR Part 50, Appendix B, Criterion V, which requires, in part, that activities affecting quality be implemented in accordance with procedures which are appropriate to the circumstance (4) Safety-related SSC issues: nine LERs dealt with cases of the failure to implement the design basis of safety-related SSC (Closed) LER 50-266/97-043-00: Inadequate TS Surveillance of Reactor Trip System interlocks. This LER described a licensee-identified condition associated with inadequate surveillance procedures for testing permissive matrix logic circuitry applicable to nuclear instrument (power range and intermediate range)
inputs for the RPS. The inadequate surveillance procedums led to the failure to perform required surveillance testing of specific contactors. This conaition was identified as the result of a licensee re-assessment of Generic Letter 96-01,
" Testing of Safety-Related Logic Circuits," applicability at Point Beach. The affected RPS functions were declared inoperable on November 15,1997, and the operating unit (Unit 1) was shut down. Corrective actions consisted of revising the applicable surveillance procedures and performing the required tests. The performance of RPS surveillance tests in a manner inappropriate to the circumstances, that is, failing to test contacts with a design basis safety-related function, was an example of an apparent violation (eel 50-266/97022-12(DRP);
50-301/97022-12(DRP)) of 10 CFR Part 50, Appendix B, Criterion V, which requires, in part, that activities affecting quality shall be implemented in accordance with procedures which are appropriate to the circumstance . - .-..- - -.- - - _ - - - - . - - . - - - . - _ - -
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(Closed) LER 50-266/97-037-00: Potential Failure of Emergency Diesel !
Generator (EDG) Load Sequence. This LER described a licensee identified 1 condition associated with the potential for safety-related loads sequencing on to the EDG powered *A" train safety-related bus to cause an undervoltage load shedding. The LER stated that this load shed would result in interruption of the
"A"t" esidual heat removal (RHR), containment spray (CS), and SW pumps for appr n.tely 25 seconds. There was no existing analysis of the effect of this delay. ~nis condition was identified as the result of a licensee self assessment of the design basis of the safety-related buses at Point Beach. Corrective actions consisted of modifications to the undervoltage load shedding circuitry for the effected "A" train buser and performance of a root cause evaluation. The "A" train EDGs were considere.1 inoperable until the required modifications were cempleted. The failu'e to provide "A" train EDGs with a demonstrated capability to support required loads under appropriate accident conditions during unit operation between the time of initial criticality and declaration of EDG inoperablity on September 3,1937, was an example of an apparent violation (eel 50-266/97022-13(DRP); 50-301/97022-13(DRP)) of TS 15.3.7.B.1.h. This TS requires that both units be shut down after seven days if a standby emergency power supply is not operable for the "A" train of ECC (Closed) LER 50-266/97-035-00: Reactor Coolant Pump (RCP) Rotor Stand Support not Seismically Adequate. This LER described a licensee-identified condition associated with an inadequately supported RCP rotor stand which was stored in an area where it could damage safety-related components during a postulated seismic event. The facility specific general design criteria requires that certain systems, including the Si and RHR systems,' be abie to withstand the loads associated with seismic events. The impact of the RCP rotor stand striking the potentially affected Si and RHR piping would have exceeded the allowable loads for that piping. The licensee concluded that this condition had existed between 1969 or 1970 and its identification on May 15,1997. This condition was identified as the result of a licensee self assessment of the design basis conformance of containment systems and structures. Corrective actions consisted of providing adequate restraints for the RCP rotor support stand and other equipment in containment. The failure to provide adequate restraints for the permanent storage of the RCP .otor stand in an area where it affected the capability of the Si and RHR systems to meet the design basis, a de facto design change, was an example of an apparent violation (eel 50-266/97022-14(DRP);
50-301/97022-14(DRP)) of 10 CFR Part 50, Appendix B, Criterion Ill, which requires, in part, that the design basis for safety . elated SSCs be correctly translated into specifications, drawings, procedures, and instructions, including those for design change (Closed) LER 50-266/97 027-00: Non Environmentally Qualified Materialin Containment Hatch Applications. This LER described a licensee-identified condition associated with the use of TEFLON material for sealing windows and in valve internals which were part of the containment personnel hatch. The TEFLON material was not qualified to withstand the postulated post-accident radiation environment to which it niight have been exposed. Damage to the TEFLON material could have affected the ability of the containment to mitigate the effects of a design basis accident. This condition was identified while evaluating operational experience at another facility. Corrective actions consisted of replacement of the non-qualified material. The failure to provide a containment capable of performing its design safety function unde, postulated accident
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i conditions while units were operati a between initial criticality and identification of this issue on May 21,1997, was an uxample of an apparent violation (eel 50-266/9702215(DRP); 50-301/9702215(DRP)) of TS 15.3.6.A.1.(a). This TS requires, in part, that if containment operability can not be maintained the {
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operating unit shall be in hot shutdown conditions within six hour (Closed) LER 50-2S6/97-018-00: Potential RHR Overpressure During Accident Conditions. This LER described a licensee-identified condition associated with the potential to overpressurize a section of RHR piping between two normally closed RHR valves inside containment. The overpressure condition would result from fluid expansion caused by higher than normal contalnment temperatures following a postulated accident. Damage to the subject piping could affect the ability of the RHR system to bring the unit to cold shutdown conditions. This was a latent design defect. Corrective actions consisted of modifications to the affected RHR system components. The failure to provide an RHR system capable of performing its design safety function under postulated accident conditions while units were operating between initial criticality and identification of this issue on April 3,1997, was an example of an apparent violation (eel 50 266/97022-16(DRP); 50-301/97022-16(DRP)) of TS 15.3.3.A.1.g. This TS requires, in part, that RHR valves and piping be capable of performing their safety functions under accident condition (Closed) LER 50-266/97-008-00: Non-seismic Ductwork Located Above Safety-related Equipment in Containment. This LER described a licensee-identified condition associated with inadequately supported ductwork that was installed over RCS pressure boundary components and other safety-related SSCs, including RHR piping. The facility specific general design criteria requires that certain systems, including the RCS and RHR systems, be able to withstand the loads associated with seismic events. The impact of the ductwork striking the potentially affected RCS and RHR piping was not evaluated when the ductwork was installed in the early 1970's or when it was modified in the 1980's. This condition was identified as the result of a licensee walkdown during the Unit 2 steam generator replacement. Corrective actions consisted of removing part of the ductwork, providing adequate restraints for part of the ductwork, and performing analysis to support the conclusion that other ductwork would not affect the operability of safety related SSCs. The failure to provide adequate restraints for the ductwork installed in an area where it affected the capability of the RCS and RHR systems to meet the design basis, a de facto design change, was an example of an apparent violation (eel 50-266/9702217(DRP);
50-301/97022-17(DRP)) of 10 CFR Part 50, Appendix B, Criterion Ill, which requires, in part, that the design basis for safety-related SSCs be correctly translated into specifications, drawhgs, procedures, arid instructions, including those for design change (Closed) LER 50-266/97-006 00: Potential Refueling Cavity Drain Failure Could Affect Accident Mitigation. This LER described a licensee-identified condition associated with the potential for a modified floor drain in the refueling cavity to fail and trap up to 46,000 gallons of water which the design basis assumed was available for core cooling in the sump recirculation mode of RHR operation. The modification to the floor drain included a non-seismic flapper valve and a reduced diameter drain nozzle. Water could have been retained in the refueling cavity behind a failed close flapper valve or a clogged drain nozzle. This condition was identified as the result of a licensee self assessment of the design basis
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conformance of containment systems and structures at Point Beach. Corrective actions included revision of the emergency operating procedures to maintain CS pump suction from the refueling water storage tank long enough to compensate for the water that could be trapped in the refueling cavity. The failure to incorporate the design basis water inventory requirements for emergency core cooling in the 1983 work package which modified the refueling cavity floor drain was an example of an apparent violation (eel 50-266/9702218(DRP);
50-301/97022-18(DRP)) of 10 CFR Part 50, Appendix B, Criterion lil, which requires, in part, that design changes be controlled to assure that the design basis is corractly translated into specifications, drawings, procedures, and instruction (Closed) LER 50-266/97-001-00: SI Delay Times Exceed Design Basis Value This LER described a licensee-identified error in the facility's accident analysi Specifically, the assumed initiation times for high head and low head SI did not include allowance for signal processing, sequencer delay time uncertainty, or the effects of degraded voltage on pump starts. The licensee had evaluations performed to demonstrate that the emergency core cooling performance criteria was achievcd when app"priate Si delay times were considered. This condition had existed from the time of initial plant licensing, but had not been identified because of inadequate system testing. This condition was identified as the result of a licensee review of Si valve stroke times. Corrective actions included revision of the Final Safety Anabis Report (FSAR) to reflect actual plant conditions, updatnig the applicabie . .fety analysis, and revising the applicable test procedure to ensure that the integrated operation of the SI system was demonstrated to satisfy design basis requirements. The failure to perform testing, from initial plant operation until this condition was identified on January 8,1997, that demonstrated the SI system would perform satisfactorily in service, that is, would initiate injection flow within the times assumed in the accident analysis, was an example of an apparent vio'ation (eel 50-266/9702219(DRP); 50-301/9702219(DRP)) of 10 CFR Part 50,4,.endix B, Criterion XI, * Test Control," wh;ch requires, in pah, that safety-related SSCs be tested in accordance with test procedures which incorporate acceptance limits contained in applicable design document (Closed) LER 50-266/96-009-00: CCW Gystem Outside Design Basis for Closed System Outside Containment. This LER described a licensee-identified condition associated with the potential for a release path from containment through the CCW system during an accident. The CCW system containment integrity relied on single safety-related containment isolation valves and maintenance of a closed system outside containment. The licensee identified that the CCW system could be open to the waste gas (WG) compressors because of make-up seal water connections (Valves WG-1030A and WG 1032A) which were active, nonsafety-related boundaries. These connections, which did not have a safety-related automatic closure feature, were inconsistent with the design basis of a closed system. This condition was identified as the result of a licensee self assessment of the 10 CFR Part 50, Appendix J, program at Point Beach. Corrective actions included maintaining the waste gas system as normally isolated from the CCW system and establishing administrative controls over the seal water make-up
supply connections when the waste gas compressors were not isolated from the CCW system. The failure to incorporate the design basis requirement that the CCW system be maintained as a closed system outside containment, that is, relying on nonsafety-related active Valves WG 1030A and WG 1032A to isolate the CCW system from the WG system, was an example of an apparent violation (eel 50-266/97022-20(DRP); 50-301/97022-20(DRP)) of 10 CFR Part 50,
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Appendix B, Criterion Ill, which requires, in part, that safety-related SSCs be controlled to assure that the design basis is correctly translated into specifications, drawings, procedures, and instructions.
I f Conclusions
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The inspectors identified 20 LERs which documented design basis conformance problems at Point Beach. Each condition was classified as an example of an apparent
- violation. The inspectors determined that the licensee had identified each of these l
conditions and had specified corrective actions which appeared appropriate to the identified condition V. Manaoement Meetinas X1 Exit Meeting Summary The ir,spectors presented the inspection results to members of licensee management at the conclusion of the inspection on February 3,1998. The licensee acknowledged the findings presented. The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified, l
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PARTIAL LIST OF PERSONS CONTACTED Wisconsin Electric Power Company (WEPCo)
S. A. Patulski, Sita Vice President A. J. Cayla, Plant Manager
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J. G. Schweitzer, Site Engineering Manage; l D. F. Johnson, Regulatory Services and Licensing Manager
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INSPECTION PROCEDURES USED IP 92700: Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor Facilities ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-266/97022-01(DRP) eel Non-exempt Power Cables do not Meet 50-301/97022-01(DRP) Appendix R Separation Criteria >
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50-266/97022-02(DRP) eel inadequately RateJ Electrical Buses Could 50-301/97022-02(DRP) Disable Switchgear and Cause Secondary Fires 50-266/97022-03(DRP) eel Electrical Short Circuits During a Control Room 50-301/97022-03(DRP) Fire Could Affect Safe Shutdown Capability 50-266/97022-04(DRP) eel Potentiai Common Mode Failure in AFW System 50-301/97022-04(DRP) Control Circuits 50-266/97022-05(DRP) eel Potential Common Mode Failure in DC Power 50-301/97022-05(DRP) Supply for AFW 50-266/97022 06(DRP) eel Nonconservative Setpoint for AFW Pump Low 50-301/97022-06(DRP) Suction Pressure Trip 50-266/97022-07(DRP) eel Redundant Safety-related Circuits in the Same 50-301/97022-07(DRP) Control Board Wireway
, 50-266/97022-08(DRP) eel Unanalyzed Service Water System Alignment i
50-301/97022-0B(DRP)
I 50-266/97022-09(DRP) eel Technical Specification Violation of Operability 50-301/97022-09(DRP) Requirement for Main Steam Line isolation 50-266/97022-10(DRP) eel Pressurizer Level Controlled Higher t%n 50-301/97022-10(DRP) Assumed in Accident Analysis 50-266/97022-11(DRP) eel Potential Si Train Failure During Filling of SI 50-301/97022-11(DRP) Accumulator 50-266/97022-12(DRP) eel inadequate Technical Specification Surveillance 50-301/97022-10(DRP) of Reactor Trip System Interlocks 50-266/97022-13(CRP) eel Potential Failure of EDG Load Sequence 50-301/97022-13(DRP)
50-266/97022-14(DRP) eel Reactor Coolant Pump Rotor Stand Support 50-301/97022-14(DRP) not Seismically Adcquate
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50-266/9702215(DRP) eel Non-Environmentally Qualified Material in 50-301/97022-15(DRP) Centainment Hatch Appliestions 50-266/9702216(DRP) eel Potential RHR Overpressure During Accident 50-301/97022-16(DRP) Conditions 50-266/97022-17(DRP) eel Non-seismic Ductwork Located Above Safety-50-301/97022-17(DRP) related Equipment in Containment 50-266/97022-18(DRP) eel Potential Refueling Cavity Drain Failure Could 50-301/97022-18(DRP) Affect Accident Mitigation 50-266/97022-19(DRP) eel Si Delay Times Exceed Design Basis Values 50-301/97022-19(DRP)
50-266/97022-20(DRP) eel CCW System Outside Design Basis for Closed 50-301/97022-20(DRP) System Outside Containment Closed 50-266/97033 LER Non-exempt Power Cables do not Meet Appendix R Separation Crneria 50 266/97032 LER inadequately Rated Electrical Buses Could Disable Switchgear and Cause Secondary Fires 50-266/97022 LER Electrical Short Circuits During a Control Room Fire Could Affect Safe Shutdown Capability 50-266/96018-21(DRS) URI Hot Smart Short Potential 50-301/96018-21(DRS) '
50-266/97041 LER Potential Common Mode Failure in AFW System Control Circuits 50-266/97036 LER Potential Common Mode Failure in DC Power Supply for AFW 50-266/57031 LER Nonconservative Setpoint for AFW Pump Low Suction Pressure Trip 50-266/96007 LER Redundant Safety-related Circuits in the Same Control Board Wireway 50-266/97030 LER Unanalyzed Service Water System Alignment 50-266/97026 LER Technical Specification Violation of Operability Requirement for Main Steam Line Isolation 50-266/97025 LER Pressurizer Level Controlled Higher than Assumed in AccMent Analysis 50-266/97009 LER Potential Si Train Failure During Filling of SI Accumulator
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50-266/97043 LER Inadequate Technical Specification Surveillance of Reactor Trip System Interlocks 50-266/97037 LER Potential Failure of Emergency Diesel Generator Load Sequence 50-266/97035 LER Reactor Coolant Pump Rotor Stand Support not Seismically Adequate 50-266/97027 LER Non-Environmentally Qualified Material in Containment Hatch Applications 50-266/97018 LER Potential RHR Overpressure During Accident Conditions 50 266/97008- LER Non-seismic Ductwork Located Above Safetprelated
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Equipment in Containment 50-266/97006 LER Potential Refueling Cavity Drain Failure Could Affect
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Accident Mitigation 50-266/97001 LER Si Delay Times Exceed Design Basis Values 50-266/96009 LER CCW System Outside Design Basis for Closed System Outside Containment
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LIST OF ACRONYMS USED IN POINT BEACH REPORTS AC Altemating Current -
AFW Auxiliary Feedwater ASME American Society of Mechanical Engineers CCW Component Cooling Water /
CFR Code of Federal Regulations CLB Current Licensing Basis CR Condition Report CS Containment Spray k EA Enforcement Action ECCS Emergency Core Cooling System eel Escalated Enforcement Action EDG Emergency Diesel Generator
,. ESF Engineered Safety Feature FSAR Final Safety Analysis Report IFl Inspection Follow-up item IP Inspection Procedure
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IR inspection Report IT In-service Test Procedure i LCO Limiting Condition for Operation i LER Licensee Event Report NCV Non-Cited Violation NDE Non-Destructive Examination
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NP Nuclear Power Department Procedures NRC Nuclear Regulatory Commission 1 Ol Operating Instruction OM Operations Manual OOS Out of-Servic OP Operating Procedure POD Prompt Operability Determination QA Quality Assurance
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RCP Reactor Coolant Pump RHR Residual Heat Removal
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RPS Reactor Protection System SSC Safety-related Structure, System, and Component
. SI Safety injection
- SW Service Water TS Technical Specification
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URI Unresolved item VIO Violation
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