ML20198Q637
| ML20198Q637 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 11/07/1997 |
| From: | Jeffery Wood CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2498, 50-346-97-201, NUDOCS 9711120290 | |
| Download: ML20198Q637 (10) | |
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Docket Number 50 346 1.icense Number NPF 3 Serial Number 2498 November 7, 1997 United States Nuclear Regulatory Commission [
Document Control Desk Washington, D.C. 20555 001
Subject:
Response to Nuc! car Regulatory Commission Inspection Report 50-346/97 201, Davis-Desse Design inspection Ladies and Get 1en:
This letter and it. tachment previde Toledo Edison's (TE) response for the Davis Besse Nuclear Power Station, Unit 1 (DilNPS), to the Nuclear Regulatory Commission's (NRC)
Inspection Report 50-346/97-201, Davis llesse Design Inspection (Log 5122), dated September 4,1997, inspection Report 50 346/97-201 presented the results of the Design Inspection conducted at DilNPS by the Office of Nuclear Reactor Regulation during the period from April 14,1997 /J through June 20,1997. During this inspection, the NRC identified three Unresolved items //
(URis) and eight inspection Follow up Items (IFis). The response to each URI and IFl and the schedule for completion of the corrective actions are presented in the attachment. ;
liased on discussion with the DBNPS Resionnt inspector on October 31,19M, the response sut,mittal date was extended to November 7,1997.
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' ' Docket Number 50 346 1.icense Number NI'l 3 Serial Number 2498 ,
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Should you have any questions or require additional infonnation, please contact Mr. James I.,
l' reels, Marager llegulatory Affairs, at (419) 321 8466.
Very truly yours, l 51 C K' ic Attachment ec: A.11. lleach, llegional Administrator, NitC Region 111 S. J. Campbell, Dil 1 NRC Senior itesident inspector A. G. Ilansen, NitC Project Manager Utility Itadiological Safety lloard f-t t
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- Docket Number 50 346 License Number NPF-3 l Serial Number 2498 !
Attachment Page 1 of 8 ,
RESPONSE TO NRC INSPECTION REPORT A0-346/97 201 FOR DAVIS IlESSE NUCl. EAR POWER STA's ION, UNIT I .
URI AND IFl CORRECTIVE ACTION COMPLETION SCllEDUI.E i
IFI 50-346/97 20101 IIPI Flow Reuulrements l'or SG Tube Hunture Accident ,
Updated Safety Analysis Report (USAR) Table 15.4.2 2," Summary of the Steam Generator !
Tube Failure Analysis," summarizes the events assumed following a Steam Generator Tube Rupture (SGTR), including the approximate time at which the liigh Pressure injection System (IIPI) wculd automatically start.
In the SGTR analysis, a double-ended break flow equivalent ofless than 435 gallons per minute (gpm), cold, was calculated by Babcock &Wilcox (currently Framatome Technologies). During power operation at pressures above the llPI shutoff head, the Make Up (MU) system would be used to ofTset the rate ofinventory loss. For the bounding double-ended tube rupture the MU l system capacity was assumed to be exceeded, with a reactor trip ensuing due to low pressure.
Following the low presstne trip, llPI would be initiated automatically, llowever, the llPI .
response time when the injection begins and the llPI injection flow rate is less demanding than for other small break LOCAs. !
To the Reactor Coolant System (RCS), this event is similar to a small break loss of coolant accident (LOCA). The tube rupture size is within the small break LOCA spectrum, but has full injection capacity always available from at least one llPI train. Since the steam genciators would remain available as heat sinks, system pressure could be reduced by operator action. In accordance with DB-OP-2000 "Reac'or Protection System, Safety Features Actuation System, Steam Feed Rupture Control System Trip or Steam Generator Tube Rupture," the operators would initiate llPI (piggybacked from Low Pressure injection (LPI)) to maximize injection capability and reduce the system pressure thereby reducing break flow, in this corSguration, at reduced pressure, the llPI system has sufficient capacity and meets the design requirements for the postulated accident. ;
Toledo Edison had requested Framatome Technologies to provide a calculation from their archives, but the calculation is not available. The values presented in USAR Table 15,4.2-2, Summary of the Steam Generator Tube Failure Analysis, will be validated by December 31, ,
1998.
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." Docket Number 50-346 l icense Number NPF-3 Serial Number 2498 Attachment Page 2 of 8 IFl 50-346/97-201-02 Safety Clan Interface at Pressure _Ga a e isolation Valves Pressure indicators (PI) 1519 and 1520 were procured as " commercial quality." The Safety Class (Q) boundary stops at the isolation valve, and the seismic bound.,ry continues to the pressure ,
indicator. The seismic boundary continues beyond the Q boundary to assure the pressure boundary is maintained durint v. Ofter a seismic event. Since the indicators are included within the seismic boundary, t'a 8ndicators, associated piping and tubing are seismically mounted, but the indicator is not seisnaally qualified. In addition, in the event that the pressure indicator were to fail to the point of not maintaining the pressure boundary, the Q isolation valve would be available to isolate the system from the failed pressure indicator. Therefore, commercial quality -
is acceptable for these applications.
The Safety Class interface at an open pressure gage isolation valve is an initial license condition depicted in the original Davis llesse Nuclear Power Station (DilNPS) Final Safety Analysis Report (FSAR). This item has been referred to Nuclear Reactor Regulation (NRR) StalTror funher evaluation. Davis isesse Nuclear Power Station completion date will be determined upon completion of the NRR review.
IFl 50-346/97 201 b2 Fyvironmental Oualification of Euulpment in the ECCS Rooms Upon identification of this condition, Potential Condition Adverse to Qtality Report (PCAQR) 97-0796 was issued to address the environmental qualification of components in the Emergency Core Cooling System (ECCS) rooms. Following evaluation, it was concluded that there was no operability concern because identical local control stations at DBNPS have been qualified for harsh environments and the raaterials commonly used for the sump system have been demonstrated to be qualified for calculated radiation levels in the ECCS rooms. The environmental qualification of this equipment will be completed by March 6,1998.
IFI 50-346/97-201-04 Itatterv Charner Surveillance Testme in USAR section 8.3.2.1.3,"llattery Chargers," the essential battery charger's design capability is stated as "Each charger is capable of supplying all steady state DC loads required under any conditions of operation while recharging the battery to a fully charged condition over a period of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from a discharged condition of 105 volts per battery." The battery charger Technical Specification (TS) Surveillance Requirement (SR) 4.8.2.3.2.c.4 states that "the battery charger will supply at least 475 amperes at a minimum of 130 volts for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />."
B During a previous NRC inspection, it was stated that the battery charger surveillance requirements' did not appear to satisfy the USAR statement of design capability (reference Electrical Distribution System Functional Inspcetion (EDSFI) Report Finding 346/92007-06).
Although Toledo Edison provided justification that the surveillance requirement was adequate to
' ' Docket Number 50 346 1.icense Number NPF 3 ;
Serial Number 2498 Attachment >
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Page 3 of 8 support the USAR statement, an inspector suggested that an enclosure be added to DB ME- .
03002," Station llattery Service and Performance Discharge Test," this ewelosure would document charger current and charge duration following a discharge of the battery. Toledo Edison added Enclosure 7 to DB ME 03002 as requested. Although implementation of :
Enclosure 7 provides useful infbnnation about battery charger performance (such as ability to produce current up to the current limit setpoint), it does not verify the USAR staternent of design capability because there are no real or simulated steady state loads on the charger during the battery recharge. ,
Toledo Edison does not take credit for Enclosure 7 of DB MF-03002 to demonstrate any charger design capabilities as described in the USAR. Therefore, the description of Enclosure 7 test results as described in the inspection report 50 346/97 201, Section El.2.3.2.a. should not be ,
misinterpreted as the method by which the Surveillance Requirement is met, nor as a method to prove the design capability as described in the USAR. Additionally, Enclosure 7 is only perfbrmed on fbur out of the six chargers per refueling period; this is acceptable as the enclosure is not used to satisfy any surveillance test or design capability assumptions.
The charger surveillance requirement is satisfied by surveillance test procedure DB ME 03003, .
" Station Battery Charger Test". This procedure utilizes a load bank to ensure that each charger ;
will supply at least 475 amperes at a minimum of 130 volts for at least eight hours. This test is ,
performed on all six chargers at an 18 month frequency in accordance with TS requirements.
Toledo Edison maintains that this surveillance test is adequate to demonstrate the USAR '
statennnt of charger design capability, and the following infonnation is presented as justification.
Toledo Edison calculation C EE 002.01-010 results show that the battery chargers are adequately sized to recharge the batteries from the condition existing at the end of a worst-case postulated Design Basis Accident scenario to their fully-charged condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> while supplying the steady-state DC loads. An engineering evaluation has been completed and indicates that the i design capability as stated in USAR section 8.3.2.1.3 is met, and that SR 4.8.2.3.2.c.4 to test the battery charger at 475 c.mperes for eight hours is sumcient to demonstrate this design capability.
! Toledo Edison concludes that the existing Technical Specification Surveillance Requirement 4.8.2.3.2.c.4 is adequate to demonstrate the charger design capability as stated in USAR Section 8.3.2.1.3. Sured!!ance test DB ME-03003 is performed on all of the essential battery chargers, -
specifically DHCl P, DBC1N, DBC1 PN (swing charger). DBC2P. DBC2N, and DHC2PN (swing charger), every 18 months.
l The results of the above mentioned engineering evaluation will be incorporated into a formal i
calculation by December 31,1997.
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l Docket Number 50 346 1.icense Number NPF 3 Serial Number 2498 Attachment 1 Page 4 of 8
- 1[I $0 346/97 201-05 Tenline ofinverter and Anociated Componesh Toledo Edison currently has no commitment to perform surveillance t r periodic testing on the regulated rectifiers, inverters, and constant voltage transformers for the essential 120VAC System. As stated in the IFI, the load on the inverter does not significantiv inange during postulated abnormal events, and therefore the ability to perfonn its safdy function is inherent in its continuous c wration. Although surveillance testing of this equir ment is not specifically required in the Technical Specification, the equipment is monitore6 for proper line up and signs of degradation by weekly surveillance test procedures DB SC 03041,"Onsite AC Bus Sources Lines Up, Available and Isolated (Modes 1,2,3 and 4)," and DB SC-03042,"Onsite AC Bus Sources Available (Modes 5 and 6)," which satisfy OBNPS Technical Specifications 4.8.2.1, "AC Distribution Operating", and 4.8.2.2,"AC Distribution Shutdown", for operability of the essential AC buses. These line-up verification tests are consistent with NUREG 1430,
" Standard Technical Specifications, Babock and Wilcox Plants" Additional monitoring occurs daily by operator rounds. The equipment is continuously monitored for abnorn al events and failures by control room annunciators and plant computer data.
The constant voltage transformers (CVTs) were added by Facility Change Request (FCR) 86-0272 to improve system reliability and are an enhancement to the original system design. These passive devices are not utilized as normal sources to the essential distribution panels and primarily exist to improve the system's ability to endure transients (via Static Transfer Switches) which could otherwise have adverse affects on an entire instrumentation channel. The inverter is always the preferred source to the distribution panels. Toledo Edison does not allow an essential distribution panel to be powered from a non inverter supplied source for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during Modes 1 through 4 by use of administrative controls. Monitoring of CVT availability is also included within the surveillance and monitoring activities described above.
Based on the above information, Toledo Edison concludes that the level of monitoring and surveillance testing on the essential 120V instrumentation AC system at DBNPS is adequate to ensure the continuous availability of the equipment and operability of the four essential 120VAC buses. Operability verification for these buses is covered by Technical Specification SR 4.8.2.1 and 4.8.2.2, and is performed by DBNPS surveillance procedures DB-SC 03041 and DB-SC-03042, respectively. These procedures include detailed verification ofinverter, rectifier, and CVT availability and status. The level of surveillance afibrded by these procedures, and supplemented by operator rounds, control room annu aciators, and plant computer inputs, continually demonstrates the readiness of the equipment to perform its intended safety function.
Docket Number 50 346 License Number NPF 3 i Serial Number 2498 :
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URI 50 346/97 201-06 USAR Discrepancien Four USAR discrepancies were noted in the inspection report. The status of resolution of these I discrepancies is as follows-l
- The USAR discrepancy involving the llPI setpoint in USAR Table 15.4.2-1," Steam i Generator Tube Failure Parameters," was licensee-identified and License Amendment Request (LAR)96-014 (Scrial Number 2441) was submitted to the NRC on April 18, 1997, to revise the Technical Specifications. The USAR will be revised upon NRC approval of the LAR.
Toledo Edison initiated USAR Change Notice (UCN)97-066, to address USAR Table 7.5-1,"Infonnation Readouts Available to the Operator for hionitoring Conditions in !
Reactor, Reactor Coolant System. Containment Vessel, ECCS, and Steam Generators." .
This UCN was approved and posted on June 22,1997. Toledo Edison initiated UCN 97-l 114 to clarify USAR Section 6.3.1.4," System Short- and Long-Term Capability,"
approval and posting of this UCN will be completed by June 30,1998.
Toledo Edison believed that the question regarding USAR Section 15.4.4.2.3.2," Double-ended hiain Steam Line lireak", as presented during the inspection. had been answered ,
and considered closed. After review of the inspection report, UCN 97-122 was initiated to clarify the USAR. Approval and posting will be completed by December 30,1997. ,
it should also be noted that DilNPS initiated an Updated FSAR Improvement Program as .
described in a letter to the NRC dated June 12,1997 (Serial Number 2468) entitled " Updated Final Safety Analysis Report initiatives and NRC Revised Enforcement Policy." This USAR Improvement Program is designed to identify and correct the type of discrepancies noted in this inspection report. This program will be completed in support of the next required revision of the USAR, which is currently scheduled for November 1998.
URI 50-346/97 20107 Reverse Flow Testine of LPI Pumn Check Valves and ilPI PumD Recirculation Ston Check Valves To address this licensee-identified item, TE initiated PCAQR 97-0529 on April 23,1997, to review the testing requirements for the LPI and IIPI systems. The potential concerns identified in the PCAQR were that the check valves, Dil81 and D1182, in the LPI injection flow path were not being reverse flow tested and the two stop check valves, llP31 and I?P32, in the llPI minimum recirculation flow pa'h were not being leak tested.
Valves llP31 and ilP32 are motor operated valves (h10V) and are included in the htOV Program set forth in Generic Letter 89-10," Safety-Related hiotor Operated Valve Testing and
Docket Number 50 346 1.icense Number NPF-3 Serial Number 2498 Attachment Page 6 of 8 Surveillance", which periodically ensures that these valves are capable of closing against the llPl pump shutofThead. During an accident, these valves would close at a pressure less than the pump shutoff head. This provides assurance that the valves would fully seat when they closed during piggy back operation.
The results of an evaluation provided to the NRC Inspection Team demonstrated that the off-site or control room doses are not significantly impacted, even if the total assumed leak rate is as high as I ppm. Although it was determined that individually leak testing IIP 31 and ilP32 was not required, it was decided that a periodic integrated system leak test would be an appropriate enhancement. Accordingly, an integrated system leak rate test was developed and approved on October 24,1997, and is scheduled to be completed by January 22,1998.
Valves Dil81 and Dil82 are in the 1,PI pump suction line from the llorated Water Storage Tank (llWST). These valves provide reverse flow protection from the Containment Emergency Sump to the llWST during the short duration of valve swapping at the initiation of containment sump recirculation. This protection is only needed if the containment pressure at the time of recirculation is higher than the llWST static pressure. The analysis of existing calculations indicated that at the time ofinitiation of recirculation the llWST static pressure is higher than the containment pressure. For this reason, reverse now testing of these valves was not considered essential. In USAll Section 6.3, reverse flow protection credit was taken Ihr Valves Dil81 and Dil82. Thercibre, it was determined that valves Dil81 and Dil82, already in the in Service Testing (IS f) program for forward How testing, will also be reverse-flow tested on a quarterly basis. The first reverse How testing was successfully completed on June 9,1997, for valve D1182 and on June 27,1997, for valve Dll81.
IFl 50-346/97-201 08 ITCS I,cakugt As stated in response to IFI 50-346/97-201-07, static pressure due to llWST water level prevents the leakage through Dil81, Dil82, Dll7A, and D11711. Thus, leak te.> ting of these valves is not considered necessary. Also, as stated above, a total leak rate of I gpm through IIP 31 and llP32 do not increase the off site doses or control room doses significantly. The Inspection lleport also discusses the performance ofleakage tests at a temperature lower than expected with accident conditions. Iteview of this issue has concluded that the temperature difference has no signincant impact on the test results. Toledo Edison recognizes the importance of keeping leakage of radioactive Guids as low as practicable. Therefore, as a result of PCAQ 97-0529, Toledo Edison will periodically perform an augmented pressure leak rate test to determine the leakage from the llPI and 1.Pl trains.
Augmented I!CCS Leakage Test Procedure, DIl-PF-04150, was developed and approved on October 24,1997. l.cak testing of both liCCS trains will be performed during the current operating cycle. This testing is scheduled to be co.npleted by January 22,1998.
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Docket Number 50 346 i License Number NPF-3 l Serial Number 2498 Attachment i Page 7 of 8 IFI 50-346M7 201-09 HWST I ow-Low Level Setnoint Ca'culations Request for Assistance (RFA) 94-0509 was not a calculation and was not intended to perform the function of a calculation it compiled the available data from previous submittala to the NRC ,
and llechtel calculations and provided the minimum acceptable actual tank level (analytical limit) at which the Safety Features Actuation System (SFAS) level 5 trip must occur, specifically,76 inches. Calculation C-lCE-048.01-004,"SFAS BWST Low Level SetpoNt,"
adds instrument string uncertainty to the 76 inch analytical limit to establish the SFAS BWST ;
low level bistable setpoint. This calculation includes the basis for instrument string accuracy and i bistable tolerance. Calculation C-NSA 049.01-004," Vortex Fonnation with ECCS Pump Suction from the BWST," issued on June 4,1997, provides the bases for the adequacy of the analytical limit. This latter calculation serves as the analysis in place of RFA 94 0509, and .
includes verification of assumptions and references for the analytical limit. Calculation C ICE-048.01-004 will be revised to change the reference for the analytical limit from RFA 94-0509 to Calculation C NSA-049.01-004 and also to include an analysis of the SFAS BWST low and high setpoints. This revision will be completed by January 30,1998.
IFl 50 346/97-201-10 llich Containment Pressure Actuation Setnoint The NRC inspection team noted inconsistencies in the documentation for the SFAS Level 3 t
containment high pressure trip. Technical Specification Table 3.3-4," Safety Features Actuation System Instrumentation Trip Setpoints," provides a trip setpoint of 18.4 psia with an Allowable Value of 18.52 psia. Bechtel letter BT-11388 addressed revising the setpoint in accordance with !
NUREG 0737,"TMI Action Plan." The letter documented an 18.24 psia setpoint and a 18.36 psia allowable value. This letter stated that the current TS values were acceptable because they compared very closely to the new calculated values. Calculation C-ICE-048.01-001, utilizing a new methodology, resulted in an instrument string uncertainty of 1.5 psi which when applied to the analytical limit would result in a setpoint of approximately 19.08 psia.
The field setpoint and TS allowable values will be revised to be in accordance with calculation C lCE-048.01-001 by December 31,1999.
URI 50-346/97-201-11 Corrective Action for HWST Level Transmitter Sunnort Corrosion The inspection team expressed a concern regarding the condition of the mounting brackets and hardware associated with the BWST level transmitters (LT1525 A, LT1525B, LT1525C and LT1525D). Although two of the transmitters located in the BWST trench were identified (LT1525B and LTl525D) as being corroded via PCAQR 94 0840, the inspection team stated that prompt corrective actions were not adequate.
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Attachment Page 8 of 8 At the time when the PCAQR was initiated in 1994, DilNPS Civil Engineering perfomied an evaluation on the condition of the level transmitters in the IlWST trench and determined that the corrosion did not impair the function of the equipment. Again on January 13,1997, the condition of the corroded brackets and hardware on the level transmitters in the llWST trench were evaluated by DilNPS Civil Engineering and it was determined that the condition did not impair the function of the equipment.
To address the inspection team's concerns, work requests were initiated to make the appropriate repairs to all four of the level transmitters. The mounting bracket and hardware for three of the level transmitters (LT1525A, LT1525C and LT1525D) were cleaned, inspected and coated to prevont further corrosion. These actions were completed on September 15,1997. The mounting bracket and hardware for level transmitter LT152511 were replaced on October 28,1997. A t review of the inspection results and maintenance activities performed on all four transmitters confinned the previous evaluations and conclusions that the corrosion did not degrade the supports to a point where there was a potential for failure.
The cause of the corrosion was moisture intrusion into the llWST trench and the adjacent shed.
Repairs have been ongoing to make thsse enclosures weather tight. The majority of the activities associated with sealing these enclosures is complete. The remainder of the repairs to make the enclosures weather tight will be completed by December 5,1997.