IR 05000456/1999013
| ML20216F748 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 09/17/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20216F747 | List: |
| References | |
| 50-456-99-13, 50-457-99-13, NUDOCS 9909240053 | |
| Download: ML20216F748 (19) | |
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l-U.S. NUCLEAR REGULATORY COMMISSION REGION lil Docket Nos: 50-456;50-457 License Nos: NPF-72; NPF-77 Report No: 50-456/99013(DRS); 50-457/99013(DRS)
Licensee: Commonwealth Edison Company Facility: Braidwood Nuclear Plant, Units 1 and 2 Location: RR #1, Box 84 Braceville, IL 60407 Dates: July 6 through August 24,1999 Inspector: Zelig Falevits Approved by: Ronald N. Gardner, Chief, Electrical Engineering Branch
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Division of Reactor Safety
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9909240053 990917 PDR G ADOCK 05000456 PDR
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U.S. NUCLEAR REGULATORY COMMISSION REGION lil Docket Nos: 50-456;50-457 License Nos: NPF-72; NPF-77
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l Report No: 50-456/99013(DRS); 50-457/99013(DRS) i Licensee: Commonwealth Edison Company Facility: Braidwood Nuclear Plant, Units 1 and 2 Location: RR #1, Box 84 Braceville, IL 60407 Dates: July 6 through August 24,1999 Inspector: Zelig Falevits Approved by: Ronald N. Gardner, Chief, Electrical Engineering Branch Division of Reactor Safety
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EXECUTIVE SUMMARY Braidwood Nuclear Plant, Units 1 and 2 NRC Inspection Report 50-456/99013(DRS); 50-457/99013(DRS)
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This was an announced inspection to review the corrective actions and engineering related activities to address the technical concerns identified during a previous NRC inspection and by the AE inspection team during the design inspection completed on April 24,199 Enaineerina
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The inspection results indicated that the corrective actions were effective in addressing the concerns identified by the AE team and steps were taken to prevent recurrence of the noted problems. (All Sections)
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Two examples of inadequate test controls were identified: (1) the licensee's testing program did not test auxiliary feedwater (AFW) diesel engine cooling system expansion tank safety-related solenoid operated relief valve; and (2) the AFW diesel engine starting circuit K11 time delay relay was inadequately tested. A non-cited violation was issued. (Sections E8.3 and E8.9)
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Two examples of inadequate design controls were identified: (1) calculation Nos. L-VA-809 and L-VA-811 did not include additional hot piping and motor heat loads in the heat capacity determinations; and (2) calculation No. CWBS-C-149 did not consider safety injection and charging pump degradation allowances in the flow determination. A non-cited violation was issued. (Sections E8.5 and E8.15)
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Two examples of inadequate procedures were identified: (1) procedure No. BwVS 0. St. 2-3 contained inadequate closure acceptance criteria for valve Nos. CV 8546 and SI 8926; and (2) procedure No. BwVS 4.6.2.2-1 had a required procedure step removed without justification. A non-cited violation was issued. (Sections E8.11 and E8.13)
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I Report Details l
Ill. Enaineerina l E8 Miscellaneous Engineering issues (92903)
E8.0 (Closed) Violation (50456/98008-02(DRS): 50457/98008-02(DRS)): The NRC 1 determined that the licensee failed to implement comprehensive and effective corrective i action to prevent recurrence of non-environmentally qualified components installed in a l harsh environmen l On April 9,1998, the licensee issued PlF A1998-01350 to document the discovery of eight examples where non-qualified safety-related components were installed in Motor Control Centers 1/2AP21E including a breaker, some thermal overload relays and a meter. The licensee performed a preliminary review of the safety-related, non-environmentally qualified items and determined that they were operable. Seven of the eight items were replaced with qualified components while one was determined to be adequately qualified. In addition, the licensee performed an extensive search of work packages which involved EQ equipment. No additional concerns were identified. A new procedure, NSWP-WM-10," Preparation of Maintenance Work Packages," Revision 0, was issued which contained improved guidance to be used for EQ application Training was provided to appropriate staff. In addition, the licensee performed a comprehensive root cause investigation and assigned coirective actions to address the findings and to prevent recurrence. The inspector reviewed the licensee's efforts to address this issue. No concerns were noted. This violation is considered close E (Closed) IFl 50456/98-201-01: The NRC identified that the potential for exceeding the AFW system discharge piping design pressure had not been evaluated when operating on minimum flow and during overspeed testing of the AFW pump diesel engine drive The team identified inconsistencies in calculations related to the piping design pressure and the maximum pressure determined during system test To resolve the inconsistencies and concerns identified by the team, the licensee issued PIF A1998-01482 to address the potential overpressure of the AFW system piping during the quarterly surveillance test. The appropriate procedures have been revised to ensure that overpressure will not occur in the future, in addition, PlF A1998-01439 was issued to address the various calculation inconsistencies regarding maximum system pressure limits. The inspectors reviewed licensees corrective actions and found them to be adequate to resolve the concems noted by the team. This item is considered close E8.2 (Closed) IFl 50456/98-201-02: The licensee's original evaluation o' NRC 'nformation Notice (IN) 90-45,"Overspeed of the Turbine-Driven Auxiliary Feedwater Pumps and Overpressurization of the Associated Piping Systems," dated July 6,1990, determined that the conditions described in the IN did not apply. However, in preparation for the NRC inspection, the licensee re-evaluated the IN and determined that it was applicable to Braidwood. The Turbine-Driven Auxiliary Feedwater pump discharge pressure could reach 600 psig at a diesel engine speed of 1925 rpm which is higher than the rated speed of 1820 rp To address this concern, the license issued PIFs A 1998-00619 and A 1998-0098 I to
document this issue and track follow up actions to formalize the maximum pressure
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calculation and determine whether future testing should be modified. Operability evaluation 98-023 was written to justify AF diesel operability. The appropriate procedures and calculations were revised to agree with the assumptions and inputs used. In addition, NTS items were generated to track required actions to resolution. No operability concerns were identified. This item is considered close E (Closed) URI 50-456/98 201-03: The team identified that the AFW diesel engine cooling system expansion tank safety-related solenoid operated valve (SOV) that opens i
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automatically on low tank pressure and the tank pressure relief cap used to prevent overpressurizing and potential failure of the diesel engine cooling system were not included in any periodic test progra To address this concern, the licensee issued engineering request ER 9800389 and revised procedure BwMP 3200-014 to incorporate the requirements to test the Jacket water expansion tank relief cap. In May 1999, the licensee attempted to perform the cap test to the revised procedure, however, the test fixture used for the first time did not fit. The licensee replaced the old cap with a new one (SIN 769A15) but this cap was also not tested. Following NRC questioning, the licensee stated that it is Braidwood's intention to fabricate the necessary test fixture to test the relief cap in the future as ,
required by the procedur l Regarding the solenoid valve, the licensee informed the team that the tank level was checked daily as part of operator rounds, and that these rounds would identify a leaking SOV. The team agreed that the operator rounds would identify a leaking SO Therefore, no additional testing of the SOV was neede l Criterion XI, " Test Control," of Appendix B to 10 CFR 50 requires that a test program be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The failure to prescribe adequate testing and test the AFW diesel engine cooling system expansion tank pressure relief cap was considered an example of a violation of 10 CFR 50, Appendix B, Criterion XI. However, this Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-456/99013-01a), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as Nuclear Tracking System (NTS) Item 456-100-98-002010 E8.4 LClosed) IFl 50-456/98-201-04: The team noted that a failure of valve 1 AF024 to open could go unnoticed by the operators since there was no valve position indication in the control room. Valve 1 AF024 is a normally closed, failed closed valve located in a common SX water return line from both AFW pumps to the SX syste PIF A1998-01337 was issued to address this concern. On February 11,1998, the licensee revised procedures BwOP AF-5," Motor Driven Auxiliary Feedwater Pump A Startup on Recirc,"and procedure BwOP AF-7," Aux Feedwater Pump B (Diesel)
Startup on Recire," and added additional operator guidance to aid the operator in identifying that AF flow to the SGs is being reduced to less than 85 GPM (minimum required AF pump flow) and ensure adequate recirculation flow discharge and/or recirc valve realignment. A procedure change was also made to monitor AFW pump recirc
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' flow rate during pump operation. The inspector reviewed the procedure changes. No concerns were note E8.5 (Closed) IFl 50 456/98-20105: The team determined that calculation L-VA-811. " Heat ,
Capacity Verification Safety injection Room 1 A/B and 2A/B," Revision 1, did not l consider the hot piping in the SI pump 1 A room during post-LOCA recirculation and the additional motor heat load when operating at design flow compared to the minimum recirculation flow condition during .the test. The calculation concluded that the maximum temperature of the Si pump 1 A room would be 106.5*F. This was based on heat loads '
obtained from tests where the pumps were running on the bypass mode at less than maximum expected design capacity and not with worst case design heat loads.
l To address this concem, the licensee issued PlF A1998-01240 on April 1,1998, to revise calculation L-VA-809 and L-VA-811 and review other cooler heat capacity calculations. As a result, the licensee identified similar problems with calculations L-VA-100,810,805 and 808. The calculations were revised to reflect the higher piping heat loads and maximum motor loads. The revised calculations showed that although the design margin was reduced, the room coolers still have adequate capacity to maintain the rooms within the required Tech Spec and design parameter Criterion Ill, " Design Control, of Appendix B to 10 CFR 50 requires that measures shall be established to assure that applicable regulatory requirements and the design j basis are correctly translated into specifications, drawings, procedures, and instructions. l The failure to include the additional hot piping and motor heat loads in various heat 1 capacity calculations was considered an example of a violation of 10 CFR 50, l Appendix B, Criterion 111. However, this Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-456/99013-02a), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as NTS Item 456-100-M402010 E8.6 (Closed) IFl 50-456/98-201-06: The team identified several discrepancies during review of calculation BRW-97-0340-E, Revision 9. The calculation determined the '
- acceptance criteria for the performance of maintenance capacity tests of the batteries for the diesel driven AF pumps.
l The calculation used 13Vdc sizing criteria for the starting motor based on tests
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performed on the AFW diesel engine at Byron in 1984. The motor supplier specified a minimum of 20Vdc for the starting motor. The team was concerned about the effects of the 13Vdc starting voltage on the qualified life of the starting motors and the increased current on the installed cables which had not been evaluated by the licensee. The licensee issued PlF A1998-01066 to document and address these concerns. The licensee also contacted the motor supplier and confirmed that the motor would perform its function at a reduced voltage of 13Vdc. The licensee determined that the AFW starting motor cables and AFW starting motors would not be adversely affected. The inspector reviewed the licensee's action to address this issue. No concerns were note This item is considered close E8.7 - (Closed) IFI,50-456/98-201-07: The team noted that the design of the AFW battery was based upon an ambient temperature of 65'F. However, the operator round sheets indicated that a room temperature of 60*F was specified as acceptable. Surveillance procedures allowed surveillances to be performed with the room temperature below the 5-l l
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design requirement of 65*F. In addition, the team noted that procedures 1BwVS 7.1.2.3.c-1, " Aux FW Diesel Prime Maker Performance Surveillance," Revision OEl, and Procedure 18wOS 7.1.2.1.a.1-2, " Diesel Driver Auxiliary FW pump monthly Surveillance," Revision 2, did not require that the batteries be charged prior to return to servic The licensee initiated PlF A1998-01043 on March 18,1998, to investigate the basis of j the 60*F criteria used during operator rounds. The licensee determined that the 60 F
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criteria was incorrect. Surveillance Procedure," Auxiliary Feed Pump Units 1 and 2 BwOSR 3.7.5.3-2," Diesel Driver Auxiliary Feedwater Pump Monthly Surveillances," and Units 1 and 2 Procedure BwOS DC-W4,"24 V DC Auxiliary Feed Pump 1/2B Battery Bank A and B weekly surveillance were revised in March 1999 to verify that room temperature is at or above battery minium design temperature of 65*F. In addition, operator Auxiliary Building weekly log sheet, Rev.16 was changed to include a check to verify that the Units 1 and 2 Procedure AF01PB room temperature was equal or greater than 65* '
PlF A1998 01263 was generated on April 2,1998, to address the concern regarding battery depletion after the AFW diesel surveillance test where the surveillance procedures did not specify that the batteries be recharged prior to return to servic l The licensee contacted the battery vendor and obtained manufacturer recommendation ,
for verifying battery capacity after the surveillance test. Nuclear Design Information l Transmittal (NDIT) BRW-DIT-98-0104 dated May 15,1998, documented the vendor's '
recommendations. Units 1 and 2 Procedures BwVS TRM 3.7.e.3," Auxiliary Feedwater Diesel Prime Mover Performance Surveillance," for Units 1 and 2 Procedure BwOSR 3.7.5.3-2 were changed to incorporate the NDIT data. This item is considered close E8.8 (Closed) URI 50-456/98-201-08: The team was concerned that replacement of Westinghouse type AR-3 relays with DC contactors in breaker closing circuits was not carried out in a timely manner to address various breaker failures. Also, a refurbishment program for the 4.16kV safety related DHP breakers was not well establishe Conditions adverse to quality of the DHP breakers that failed had not been promptly corrected and the causes of the breaker failures had not been determine Licensee investigation concluded that Braidwood established a low priority for the proposed design change to replace the AR-3 relays with DC contactors since it was found to be technically adequate with both the existing configuration and the proposed modification, investigation revealed that the AR-3 relay contact failures at Byron were due to malfunction of the switchgear breaker (i.e., the breaker did not close due to a mechanical binding problem) and not directly related to the design of the AR-3 relay itself. To date, Braidwood has issued five design packages for installation of the AR-3 relays. The remaining 10 design packages are scheduled to be issued by December 2000. Review of failure history at Braidwood did not show that AR-3 relays had failed in a manner similar to failures noted at Byro The licensee presented the inspector with a refurbishment schedule for the 4.16KV safety related DHP breakers. Approximately 45 of the 79 safety related DHP breaker have been refurbl~shed. No generic problems were noted with the refurbished breaker This item is considered close E8.9 (Closed) URI 50-456/98-201-09: The team identified that time delay relay Kil used in the DDAFW pump start circuit, which provides a 10-second time delay of the low lube oil pressure trip function to allow the engine to start and establish oil pressure, was not adequately tested or periodically calibrated. The team was concerned that if the time delay relay drafted outside of the required 10-second time delay setting due to the lack of calibration and performance trending, the DDAFW pump might not start when required. The test procedure did not include the testing of the Kil relay and its contact 3-5 to demonstrate that the contact will defeat the 10 psig low oil pressure trip during engine star To address this concern the licensee issued PIF A1998-01073. Surveillance procedure BwHS 4002-091, " Time Delay Relay Surveillance," was revised to include relay K11 and PM data sheets were generated to ensure that the relay will be calibrated and tested periodically. The licensee evaluated the existing testing of the AF diesel engine and concluded that the existing testing was sufficient to meet the diesel driver AF pump Technical Specification requirements. The licensee stated that the K11 relay function was being verified through a monthly run of the diesel driven pump. However, conformance to the 10 second time delay required by the design drawings to ensure that the Kil time delay setting had not drifted was not being performed by the license l l
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Criterion XI," Test Control," of Appendix B to 10 CFR 50 requires that a test program be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. The failure to adequately test and calibrate the K11 time delay relay used in the diesel engine starting circuitry for the diesel drive Auxiliary Feedwater (AF) pumps was considered an example of violation of the requirements of 10 CFR 50, Appendix B, Criterion XI. However, this Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-456/99013-01b),
consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensees corrective action program as NTS Item 456-100-98-0020109.
E8.10 (Closed)IFl 50-456/98-201-10: The team identified that the sensing line for CST level transmitter ILT-CD051 was not completely heat traced. Contrary to P&lD M-39, sheet 1, about 3 feet of the sensing line was left without heat tracin PlF A1998-01362 was generated to address this discrepancy. The licensee determined that heat tracing was not required and that Armaflex insulation was adequate freeze protection for this line in July 1998, the licensee installed the Armaflex insulation using WRs 980051032 and 980051026. The design drawings were revised to eliminate the heat trace. This item is considered closed.
E8.11 (Closed) URI S0-456/98 201-11: The team noted that the licensee's evaluation of Information Notico (IN) 91-56, documented as NTS item 456-103-91-05600, concluded that the issues identified in the IN did not exist at Braidwood since the RWST was vented to the auxiliary building filtered ventilation system. Information Notice (IN) 91-56,
" Potential Radioactive Leakage to Tanks Vented to Atmosphere," identified potential problems resulting from leakage of post-LOCA recirculating containment sump water through isolation valves to the RWST and out the tank vent, thus contributing to off-site and control room dose. The valves preventing leakage to the RWST are the CV, RHR, and SI pump suction check valves CV8546, RH8958A and B, and Sl8926 respechvel ,
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If the associated motor-operated valves were open because of the valve operating sequence specified in procedures or in the event of a single failure to close, these check valves would have to prevent back flow to the RWST. The Si minimum flow isolation valves, Sl8814 and Sl8920, also prevent leakage of containment sump water to the RWST. The team determined that these valves were tested for closure but were not leakage teste The need to test closure of check valves was identified in September 1996, as documented in PIF 456-201-96-1953. The team reviewed the October 1997 closure tests performed for CV8546 and Sl8926 that were conducted in accordance with procedure BwVS 0.5.2.SI.2-3, " Safety Injection System Check Valve Stroke Test,"
Revision 9. Each check valve closure was tested by pressurizing the appropriate pump (CV or SI) suction piping with RHR pump flow and verifying that the pressure in the '
suction piping was above a certain acceptance criterion. The team determined that the closure test acceptance criterion could be met even if the check valve was not fully closed. The licensee agreed with the team, and initiated PIF A1998-01475 to determine the actions necessary to properly test closure of these valves. The licensee determined the check valves were operable on the basis of previous valve testing, inspection, and valve exercising. Testing of closure of check valves CV8546 and Sl8926 did not {
demonstrate that they will perform satisfactorily in service. To address this concern, the licensee initiated PIF A1998-01475. Procedure 5.5.8.SI.4 was issued to incorporate improved testing of the CV and Si valve !
I Criterion V, " Instructions, Procedures, and Drawings," of Appendix B to 10 CFR 50 l requires that activities affecting quality shall be prescribed by documented instructions, j procedures, and drawings of a type appropriate to the circumstances. Inadequate acceptance criteria in test procedures is considered an example of violation of 10 CFR 50, Appendix B, Criterion V. However, this Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-456/99013-03a), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as NTS Item 456-100-98-002011 E8.12 (Closed) IFl 50 456/98-201 12: Page E.77-3 of Revision 6 of the UFSAR stated that l the RWST and associated piping were excluded from the leak reduction program for i systems that could potentially contain radioactive fluids after a LOCA and stated that the l SER supplement 1 issued by the NRC confirmed such exclusion on page 9- Therefore, leakage testing of the check valves and minimum flow isolation valves between the RWST and the ECCS pumps was not required. The team believed that the i NRC SER allowed exclusion of RWST and its associated piping from the leak reduction 1 program because the UFSAR stated that highly contaminated water is prevented from entering the RWST. This issue of leakage testing was referred to the NRC staff for j further review. Subsequent NRR review indicated that leakage testing was not require This item is considered closed.
l E8.13 (Closed)IFl 50 456/98-201-13: Engineering Request (ER) 9501896 concluded that leakage measurement of reactor coolant system isolation valves by opening valves 1/2RH028A/B,1/2Sl044,1/2Sl053,1Sl052, or 2Sl050 in the vent / drain lines and collecting the leakage in a graduated cylinder, must include the use of a temporary 3/8 inch inside diameter orifice in the line to restrict the loss of reactor coolant in order for the evolution to be bounded by the existing UFSAR design basis analysis.
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Procedure BwVS 4.6.2.2-1, " Reactor Coolant System Pressure Isolation Valve Leakage
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Surveillance," Revision 14, did not require orifices to be installed when using valves 1/2RH028A/B for leakage measurements. The licensee stated that Revision 9 of the procedure added steps to install orifices on all the vent / drain valves used for leakage measurement and the steps to install the orifices for the RH028 valves were removed in Revision 12. No justification for removal of the steps was availabl The licensee determined that past tests without the orifice did not present a condition outside the design basis and initiated PlF A1998-01231, which proposed that the revision history of BwVS 4.6.2.2-1 be further researched to determine why the requirement for the orifice was removed, to evaluate the inclusion of the requirement, and to revise procedure BwVS 4.6.2.2-1. The licensee reviewed past revisions and could not determine why the requirement was removed. The procedure was vised and the steps were added back to procedure BwVS 4.6.2.2-1 as originally require Criterion V," Instructions, Procedures, and Drawings," of Appendix B to 10 CFR 50 requires that activities affecting quality shall be prescribed by documented instructions, procedures, and drawings of a type appropriate to the circumstances. Removal of required procedural steps in a surveillance procedure is considered an example of violation of 10 CFR 50, Appendix Violation (NCV 50-456/99013-03b), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as NTS ltem 456-100-98-002011 E8.14 (Closed) IFl 50-456/98-201-14: During a self-assessment in preparation for the AE inspection, the licensee had initiated PIF A1998-00670, "Si Accumulator piping safety / seismic classification," dated February 19,1998, to identify that the remotely operated valves Sl8878A-D, S18877A-D, Sl8875A-D, and S10943 used to fill, drain, vent, I and pressurize the Sl accumulators were not in the IST program and could not be relied upon to close in the event of a LOCA during these activities. These valves also form a boundary between Category IB and Category llD piping. The corrective actions for the PIF were to determine if a Limiting Condition of Operation (LCO) should be entered during the above evolutions, determine the appropriate testing requirements for the valves involved, and sample other systems to determine if other similar conditions existe The licensee determined that the subject Si valves were not within the scope of the IST program but periodic testing was a good practice. Procedure BwOP SI-1 was created to test the valves every three years, in addition, BwOP SI-9 was revised to require LCOAR entry whenever 1/2 SI 0943 and 1/2 SI 8875A or 1/2 SI 88758 or 1/2 SI 8875C or 1/2 SI 8875D are open concurrently in modes,1,2, and 3 with RCS pressure greater than 1000 psig. Also, a sample review identified other valves in the PS system that required periodic testing. All licensee action items had been completed under NTS items 456-201-98-CAOS-0046401 through 456-201-98-CAOS-00046411. This item is considered close J E8.15 (Closed)IFl 50 456/98 201-15: In letters to the NRC dated February 21,1989, and June 30,1989, the licensee responded to NRC Bulletin 88-04 and stated that the safety injection, charging (high head safety injection), and auxiliary feedwater pumps were not susceptible to the potential for dead heading due to pump-to pump interactio However, the team noted that calculation CWBS-C-149, dated June 16,1988, which caiculated weaker Si and charging pump flows did not take into consideration the pump degradation allowed by Section XI of the ASME Code as required by the NRC bulleti l 9 !
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The licensee issued PIF A1998-1259 to re-perform the calculation. With regard to the AFW pumps, the licensee stated that there were no scenarios where both AFW pumps are run aligned to minimum flow. However, the NRC pointed out that emergency procedure 18wEP- O allowed operation of both AFW pumps at minimum flow to control steam generator level. The licensee issued PIF A1998-01395 to include the AFW and charging pumps in the minimum pump flow evaluation The licensee re-performed calculations BRW-98-0384-M, BRW-98-0394-M and BRW-98-0535-M for the SI, AF and CV pumps to address the deficiencies noted in the original response to Bulletin 88-24. Calculation results indicated that for the Unit 1 Si and CV pumps the strong pump could prevent the weak pump from providing the minimum flow required for pump operation in recirculation mode. ERs 9801544 and l 9801436 were generated to develop actions needed to preclude deadheading of the l weak pump. The licensee also determined that no action was required for Unit 2 Si and j CV pumps to prevent pump deadhead. With regard to the AF pump the licensee i determined that the current configuration was acceptable with minimum pump flows of 61 gpm for the duration not to exceed 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per mont Criterion ill, " Design Control," of Appendix B to 10 CFR 50 requires that measures shall ,
be established to assure that applicable regulatory requirements and the design basis '
are correctly translated into specifications, drawings, procedures, and instructions. The failure to ensure that calculations are correctly performed and translated into surveillance procedures is considered an example of a violation of 10 CFR 50, Appendix B, Criterion Ill. However, this Severity Level IV violation is being treated as a j Non-Cited Violation (NCV 50-456/99013-02b), consistent Appendix C of the NRC '
Enforcement Policy. This violation is in the licensee's corrective action program as NTS Item 456-100-98-002011 i E8.16 (Closed) IFl 50-456/98-201-16: The team noted that the relief valv'e provided for the non safety-related sample cooler panel 1PS29J was incorrectly installed so that closure of the CCW supply and return isolation valves CC9477A and CC9448B for the cooler panel would have isolated the cooler from the relief valve. UFSAR Sections 9.2.2.2. and 9.2.2.4.1 stated that relief valves were installed in the CCW system to relieve the volumetric expansion which would occur in the cooled components if the CCW were isolated and heat added to the isolated portion of the CCW syste The licensee issued PlF A1998-01215 to document this issue. Plant Design Changes D20-1/2-99-307 were initiated to remove isolation valves 1/2CC9447A thus ensuring that the relief valve can not be isolated from the sample panel skid. The isolation function of valves 1/2CC9447A will be performed by isolation valves 1/2CC9448A. In the interim, the licensee placed caution cards on valves 1/2CC9448B to provide an adequate relief path for the relief valves. This item is considered close E8.17 (Closed) IFl 50-456/98-201-17: Information Notice 91-45,"Possible Malfunction of Westinghouse ARD, BFD, and NBFD Relays, and A200 DC and DPC 250 Magnetic Contactors," dated July 5,1991, and Supplement 1, dated July 29,1994, identified that these relays might malfunction due to an epoxy compound used in coils becoming semi-fluid when the coil was energized for extended periods. The licensee evaluated this notice and concluded that no NBFD relays were installed at Braidwood, in response to the team's questions, the licensee determined that NBFD relays were installed as "Y"
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relays on all medium voltage switchgear. The licensee issued PIF A1908-01483 to ,
document and investigate this discrepanc l Licensee investigation determined that the "Y" relays used in the anti-purtping circuitry of the Westinghouse DHP series breakers are normally de-energized ann are energized only for a short period of time following breaker closure. IN 91-45 primaily discusses j
device failures, including NRED relays, in continuously energized applimtion l Therefore, IN 91-45 was determined not to be directly applicable to the 3raidwood l NBFD relay application. The licensee plans to replace the NBFD relayr during Westinghouse breaker refurbishments. This item is considered close E8.18 (Closed) URI 50 456/98-201-18: The team noted that a design deficiency in complying with electrical separation requirements had not been promptly corrected. Note (t) on 1
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UFSAR Table 8.3-5 states that the turbine bearing oil pump (TBOP) is powered from the class 1E 480-V switchgear, however, it automatically trips on a safety infection signal concurrent with a loss of offsite power. The team noted that upon receipt of a SI signal the motor is stripped from the bus, but was designed to be automatically loaded back on to the bus after the load shedding had been accomplished. The licenses had identified this condition in 1992 during preparation for the Electrical Distribution Satety Functional Inspection (EDSFI), and initiated NTS item 925-20-92-00500 on September 14,1992, and ER 9500634 to track modification to the control circuit This design dd not meet the ,
intent of the UFSAR commitments relative to electrical separation requirements in l Regulatory Guide 1.75, Revision 2, since non safety-related loads are stripped from safety-related buses to achieve electrical separation. Automatically reloading the motor on to the bus would present a condition where a single failure of a circuit bmaker to operate could allow a fault in the non safety-related motor to propagate to safety-related equipment. In 1996, the licensee issued PIF 456-201-96-2708 to document this condition and performed an operability assessment. The operability determination stated that although the design logic of the TBOP is contrary to the UFSAR tad SER commitments, the condition has no safety significance because it will not degrade the performance of the Class IE system below an acceptable level. However, the operability determination further stated that the corrective action required for this issue was to perform the design change to provide the required tripping of the TBOP feed b'eaker with a SI/ LOOP signa Exempt Change E20-1-97-277 was issued on April 7,1998, and Exempt Change E20-2-97-277 was issued on July 24,1988, to revise the turbine bearing oil pump trip control circuitry to prevent automatic reloading on the bus after the pump was strippe The licensee subsequently installed the modifications on Unit 1 in October 1998, and on Unit 2 in May 199 E8.19 (Closed) IFl 50-456/98 201-19: The team noted that the battery charger sizing calculations did not include the inverter loads during normal or accident conditions. The team determined that the inverter could be powered by DC under certain conditions amd not by the normal AC source as designed. The team was also concerned that the output voltage of the charger could be high enough to become the primary supply to the 120V DC inverters. No engineering limits were placed on the charger output voltage settings to prevent this from occurring. The licensee performed a review of daily operator rounds sheets and determined that the inverters were being powered from the normal power AC source. The licensee could not confirm, however, by using the
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analytical model, that the inverters would be powered from the AC power supply under all condition l The licensee issued PIFs A1998-01469 and PlF A1998-01468 to address these concerns. The licensee revised calculation BYR 97-205/BRW 97-0383-E,"125VDC Battery Charger Sizing Calculation," to explicitly indicate that the inverters have been j considered loads on the battery chargers and provided supporting references. The l
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. licensee als9 verified by walkdown and observations that the inverters remained on the AC feed throughout the normal range of switchyard voltages. Revision to the calculation also addressed the concern that the inverters might become a load on the battery chargers when the batteries are equalized by mandating that the crosstie configuration not be utilized during battery equalization or when recharging a completely discharged
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battery in order to preclude overloading of the chargers. This item is considered close l E8.20 (Closed) IFl 50 456/98-201-20: UFSAR Section 8.3.1.2 states that the electrical system is designed to prevent automatic load shedding of the emergency power buses once the onsite sources are supplying power to all sequenced loads on the busesa' fter a loss of power. The team noted that the load shed interlock feature uses the "b" contact of the respective diesel generator breaker. This interlock defeats the load shedding feature while the loads are being fed from the onsite source. The load shed feature is reinstated when the diesel generator breaker is open and the loads are fed from the offsite source. The function of the "b" contact was not tested. The licensee stated that this potential weakness in the testing program had been identified in 1993, and again as part of the GL 96-01 review, and would be addressed during that revie To address this concern the licensee revised surveillance procedures 1/2BwVSR 3.3.5.3-141 and 1/2BwVSR 3.3.5.2-142 on March 29,1999, and added steps to verify the " blocking" function of the "DG" output "b" contacts to verify DG breaker UV interlock function. This item is considered close E8.21 (Closed) URI 50-456/98-201-21: The vendor manual for the safety-related 125V DC battery chargers, L-0520, " Instruction Manual, Three Phase Thyristor Controlled," dated i
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June 29,1979, stated in Section 6.6 that "As a result of qualification to IEEE-323, the following maintenance program must be adhered to: ... circuit breakers must be operated at least every 6 months. The breakers should have voltage applied and be operated under lead." The team determined that the circuit breakers supplied with the battery chargers were not tested in accordance with the vendor manual recommendations. The licensee stated that the breakers were cycled without load during the performance of BwHS 4009-003, " Clean and inspect Station Battery Chargers," Revision OE1, and that this cycling met the intent of the vendor manua Procedure BwHS 4009-003 did not, however, contain any specific steps requiring that these circuit breakers be cycled. The team noted that the lack of 125V DC breaker testing was also considered a testing weakness during the EDSFl conducted in 199 The inspector, during follow up inspection, examined licensee maintenance practices on the 125 VDC battery chargers breakers. In addition, failure history did not show any past failures of these breakers to function on demand or during testing. The licensee stated that the breakers are cycled several times during performance of BwHS 4009-003. This cycling appears to meet the intent of the vendor manual. This item is considered close I
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l E8.23 (Open) URI 50 456/98 201-23: Calculation SITH-1, " Refueling Water Storage Tank i
(RWST) Level Set points," Revision 4, contained a data sheet and acceptance critena l that were used during simulator drills to verify that the switchover of Si from injection to I recirculation could be performed before the RWST reached a level which could cause adverse suction conditions for the ECCS pumps. The data sheet was based on flow rate calculations that considered pump flows and backflow from the RWST directly to the containment sumps when the RHR pump suction was simultaneously open to the RWST and the sumps. The backflow was a major factor in the development of the data sheet since it was about 50 percent or more of the total flow from the RWST depending on valve alignment. The calculation stated that this backflow rate was a conservative ;
engineering judgement that did not require an approved calculation dJe to the level of I conservatism used but provided no basis for this judgement in the calculation. In I response to the team's questions, the licensee retrieved the basis for the engineering j judgement which was an informal bounding engineering analysis that included .
developing a piping model and checking the model against Westinghouse results for backflow during a large break LOCA. The team's review of this analysis determined that the flow rates calculated were not conservative because it did not maximize flow from I the RWS The licensee performed a preliminary re-evaluation of the calculation considering other conservatisms in the assumptions used for pump flows and valve opening times, and concluded that the acceptance criteria on the data sheet used to verify adequate l switchover time were still acceptable. The licensee initiated PIF A1998-01440 to revise the calculation. The licensee revised calculation SITH-1 on June 10,1999, to better document engineering judgement and assumptions and to demonstrate that the assumptions made were correct. This item is considered close E8.24 (Closed) URI 50-456/98-201-24: The protective relays mounted within the switchgear for EDGs 20E-1-4020A and 20E-1-4021 A were type SA-1 generator differential relays used for the differential current relay circuit that tripped the EDG circuit breaker on generator differential overcurrent. The team questioned whether the electrolytic capacitors on the relays had a replacement schedule, because they were known to degrade over time and adversely affect circuit performanc The team noted that product manual LO297 "4160-6900 V Switchgear and Bus Duct Volume 2," dated November 21,1996 (Braidwood number BwAP 1340-ST3, Revision 0)
required that all relays should be checked once a year to catch the electronic component failures which occur on a random basis and that the tantalum capacitors may have a common mode failure characteristic and should be checked visually for symptoms of electrolyte leakage every year and replaced if necessary. The manual also stated that the tantalum capacitors should be replaced every ten year The SA-1 relays were checked for function, operation, and calibration every 36 months in accordance with BwHS 4002-066, " Periodic Protective Relay Calibration," Revision No routine replacement program had been established for the tantalum capacitors. The licensee stated that no adverse trends in the performance of tantalum capacitors, transistors or potentiometers had been noted during periodic maintenance of these relays, and issued PIF A1998-01346 to evaluate the SA-1 relay maintenance requirements. Activities affecting the performance of the SA-1 relays had not been prescribed in the procedur .
ER 9801226 was issued to evaluate the manufacturers replacement recommendations for SA-1, KF, SSV-T, and SSC-T relays. The evaluation concluded that various components should be checked and/or replaced periodically. In addition, the licensee determined that the latest vendor's instructions for the SA-1 relays (41-438.11B) no longer recommended that the capacitors be replaced every 10 years. The licensee generated the appropriate predefines to ensure that the ER recommendations are implemented. This item is considered close E8.25 (Closed) URI 50-456/98 201-25: During a LOCA, the draw down of water from the I RWST would result in a partial vacuum in the tank air space due to the arrangement of the vent piping. Check valve 1Sl8969F in the common line connectinn the " low" side of the level transmitters is intended to prevent air from the RWST tunnel teaking through to the level transmitter reference legs. The check valve is normally closed and opens intermittently to drain condensate that may have collected in the common level transmitter reference leg The licensee concluded that the problem postulated would not be a concern since leakage through check valve 1/2 Sl8969F will not adversely affect the accuracy of the RWST level instrumentation and the level sensed by the transmitter will still be the true water level in the tank. The inspector discussed this issue with the licensee and examined the licensee's conclusions. This item is considered close l IV. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee staff in an exit meeting on August 24,1999. The inspectors noted that no documents provided during the inspection were identified as proprietary. The licensee acknowledged the information presented and agreed that no proprietary information was provided to the inspector !
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PARTIAL LIST OF PERSONS CONTACTED
L icensee i A. Bejantour, Design Engineer ,
l M. Cassidy, Regulatory Assurance - NRC Coordinator l
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F. Lentine, Design Engineering Manager !
T. Luke, Site Engineering Manager J, Matthews, System Engineer '
l B. Viehl, Senior Staff Engineer
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INSPECTION PROCEDURES USED IP 92903: Follow-up - Engineering l
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ITEMS OPENED, CLOSED, AND DISCUSSED l l
Ooened l 50-456/99013-01a NCV inadequate AFW Diesel Engine Cooling Test 50-456/99013-01b NCV Inadequate AFW Diesel Engine Time Delay Relay {
Test )
i 50-456/99013-02a NCV Inadequate Heat Capacity Calculation
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50-456/99013-02b NCV inadequate Design Controls 50-456/99013-03a NCV Inadequate Test Procedure Acceptance Criteria 50-456/99013-03b NCV Inadequate Surveillance Procedure Closed 50-456/99013-01a NCV Inadequate AFW Diesel Engine Cooling Test
' 50-456/99013-01b NCV Inadequate AFW Diesel Engine Time Delay Relay Test 50-456/99013 02a NCV inadequate Heat Capacity Calculation 50-456/99013-02b NCV Inadequate Design Controls 50-456/99013-03a NCV inadequate Test Procedure Acceptance Criteria 50-456/99013-03b NCV inadequate Surveillance Procedure 50-456/98-201-01 IFl AFW Piping Design Pressure 50-456/98-201-02 IFl AFW Pump Overspeed Testing 50-456/98-201-03 URI Diesel Engine Pressure Relief Cap Testing 50-456/98-201-04 IFl AFW Minimum Flo'n 50-456/98-201-05 IFl Room Heat Capacity Verification Calculations 50-456/98-201-06 IFl DDAFW Battery Calculation Discrepancies 50-456/98-201-07 IFl Electrical Test Procedure Discrepancies 50-456/98-201-08 URI ESF Circuit Breaker Failures 50-456/98-201-09 URI Testing of K11 Relay 50-456/98-201-10 IFl Instrument Tubing Heat Tracing
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o 50-456/98-201-11 URI Check Valve Reverse Flow Testing 50-456/98-201-12 IFl ECCS Leakage Through Valves 50-456/98-201-13 IFl Orifice Used in Test Procedure 50-456/98-201-14 IFl St Accumulator Operations 50-456/98-201-15 IFl ECCS and AFW Pump-to-Pump Interaction 50-456/98-201-16 IFl CCW Relief Valve Location 50-456/98-201-17 IFl Evaluation of NBFD Relays 50-456/98-201-18 URI Automatic Loading of Turbine Bearing Oil Pump 50-456/98-201-19 IFl Inverter Power Supply 50-456/98-201-20 IFl EDG Circuit Breaker Testing 50-456/98-201-21 URI Battery Charger Circuit Breaker Testing 50-456/98-201-23 URI Verification of Design input Analysis 50-456/98-201-24 URI EDG Relay Maintenance 50-456/98-201-25 URI RWST Level Instrument Check Valve 50-456/457/98008-02(DRS) VIO Inadequate corrective action to prevent recurrence of installation of non-environmentally qualified componen Discussed None
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I LIST OF ACRONYMS USED AFW Auxiliary Feedwater BwAP Braidwood Administrative Procedure BwMP Braidwood Maintenance Procedure CFR Code of Federal Regulations DRS Division of Reactor Safety ER Engineering Request IN information Notice NCV Non-cited Violation NDIT Nuclear Design Information Transmittal NEP Nuclear Engineering Procedure NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSP Nuclear Station Procedure '
NTS Nuclear Tracking System PIF Problem Identification Form UFSAR Updated Final Safety Analysis Report t
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