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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20212J0501999-09-21021 September 1999 Safety Evaluation Re Licensee Implementation Program to Resolve USI A-46 at Plant,Per GL 87-02,Suppl 1 ML20210L8661999-08-0202 August 1999 Safety Evaluation Accepting License 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20195K1481999-06-16016 June 1999 Safety Evaluation Authorizing Relief Request RV-23A for Duration of Current 10 Yr IST Interval on Basis That Compliance with Code Requirements Would Result in Hardship Without Compensating Increase in Level of Quality & Safety ML20205Q5291999-04-16016 April 1999 SER Concluding That Quad Cities Nuclear Power Station,Unit 1,can Be Safely Operated for Next Fuel Cycle with Weld O2BS-F4 in Current Condition Because Structural Integrity of Weld Will Be Maintained ML20205J6011999-04-0707 April 1999 Safety Evaluation Accepting Proposed Merger of Calenergy Co, Inc & Midamerican Holdings Co for Quad Cities Nuclear Power Station,Units 1 & 2 ML20196D9651998-11-30030 November 1998 Safety Evaluation Supporting Relief Requests CR-21 & CR-24, Respectively.Relief Request CR-23,proposed Alternative May Be Authorized,Per 10CFR50.55a & Relief Request CR-22 Was Withdrawn by Licensee ML20196A9761998-11-20020 November 1998 Safety Evaluation Re Licensee 180-day Response to GL 95-07, Thermal Binding of Safety-Related Power-operated Gate Valves ML20151T2711998-09-0404 September 1998 Safety Evaluation Accepting Licensee Response to NRC Bulletin 95-002 ML20216F0221998-03-0606 March 1998 Safety Evaluation Accepting Request Re Temporary Use of Current Procedure for Containment Repair & Replacement Activities at Plant ML20197B9171997-07-23023 July 1997 Safety Evaluation Re Concrete Expansion Anchor Safety Factors for High Energy Line Break Restraints ML20141E5091997-05-16016 May 1997 Safety Evaluation Supporting TR EMF-96-051(P), Application of Anfb Critical Power Correlation to Coresident GE Fuel for Plant,Unit 2 Cycle 15 ML20137G6071997-03-13013 March 1997 Safety Evaluation Supporting Proposed Changes to TS & Bases Ceco ML20134H7601997-02-0707 February 1997 Safety Evaluation Approving Rev 65c of Ceco QA TR CE-1-A ML20149F4151994-08-0404 August 1994 Safety Evaluation Concluding That Unit 1 Can Be Safely Operated During Next Operating Cycle (Cycle 14) ML20058L2711993-12-0808 December 1993 Safety Evaluation Finding Overlay Repair of Weld 02C-F7 Acceptable & in Conformance W/Gl 88-01.Plant May Be Returned to Safe Operation ML20056C4601993-06-17017 June 1993 Safety Evaluation Accepting Proposed Repair of Weld in Recirculation Piping Sys for One Cycle of Operation ML20128F9731993-02-10010 February 1993 Safety Evaluation Granting Licensee 910930 Request Not to Perform Code Exam on 100% of Attachment Welds on Stabilizer Brackets to Reactor Vessel Under 10CFR50.55(a)(3)(ii) ML20055F9221990-07-17017 July 1990 Safety Evaluation Supporting Util Responses to NRC Bulletin 88-010 Re Molded Case Circuit Breaker Replacement ML20248J2431989-10-0303 October 1989 Safety Evaluation Accepting Util 880122,0601,0714 & 0816 Submittals Re Insp Results,Mitigation,Flaw Evaluations & Overlay Repairs of Welds Susceptible to IGSCC to Support Operation of Unit 2,for Another 18-month Fuel Cycle ML20246K1611989-08-24024 August 1989 Revised SER Supporting Amends 112 & 108 to Licenses DPR-29 & DPR-30,respectively,changing Setpoints of Main Steam Line Radiation Monitors & Correcting Typos in Tech Specs ML20248B8911989-06-0606 June 1989 Safety Evaluation Concluding That IGSCC Insp Scope for Class 1 Piping Meets NRC Requirements & Guidelines of Generic Ltr 84-11 ML20151X3431988-08-16016 August 1988 SER Accepting Basis & Findings That Util post-accident Monitoring Instrumentation Meets Guidelines of Reg Guide 1.97 Except for Variable Neutron Flux Instrumentation ML20151M6901988-07-21021 July 1988 Revised Safety Evaluation Supporting Exemption Requests from Regulatory Requirements of 10CFR50,App R,Section Iii.G ML20195E2091988-06-0909 June 1988 Safeguards Evaluation Rept Supporting Amends 108 & 103 to Licenses DPR-29 & DPR-30,respectively ML20151U1201988-04-20020 April 1988 Revised Safety Evaluation Accepting Util Interim Compensatory Measures & Request for Exemption from 10CFR50, App R,Section Iii.G Requirement Re Hot Shutdown Repair for Fire Event in Plant ML20149M5301987-12-11011 December 1987 Marked-up Safety Evaluation Supporting Request for Exemptions from App R ML20236W4851987-12-0101 December 1987 Safety Evaluation Accepting Proposed Approaches for Resolving fire-related Concerns,Including Spurious Operations,High Impedance Faults & Electrical Isolation Deficiency.Granting of Exemption Requests Recommended ML20235S8541987-10-0202 October 1987 Safety Evaluation Supporting Interim Approval of Rev 3 to Process Control Program for Plant ML20237H7061987-08-19019 August 1987 SER Supporting Util Response to Item 2.1 (Part 1) of Generic Ltr 83-28 Re Equipment Classification.Licensee Statements Confirm Program Exists for Identifying,Classifying & Treating Components as safety-related.Program Acceptable ML20236H1341987-07-27027 July 1987 Safety Evaluation Re Acceptance of Updated Rev 11 to Offsite Dose Calculation Manual ML20205H1351987-03-23023 March 1987 Safety Evaluation Re Insps for & Repairs of Igscc.Facility Can Be Safely Operated for One 18-month Fuel Cycle in Present Configuration ML20214X1111986-11-26026 November 1986 Safety Evaluation Supporting Util Analytical Methods Used to Evaluate Stresses of Critical Components for Vacuum Breaker Integrity Re Mark I Containment Program ML20214Q3851986-11-17017 November 1986 Safety Evaluation Re Insp & Repair of Reactor Coolant Piping Sys ML20141D2291986-03-31031 March 1986 Safety Evaluation Granting Util Request for Relief from Certain Requirements of Section XI of ASME Code Re Inservice Insp for Second 10-yr Interval ML20141P0491986-03-13013 March 1986 Safety Evaluation Supporting Licensee 831105 & 851219 Responses to Generic Ltr 83-28,Item 1.2, Post-Trip Review (Data & Info Capability) ML20137A3931986-01-0707 January 1986 Safety Evaluation Supporting Reactor Coolant Piping Sys IGSCC Insp & Repair Per Generic Ltr 84-11 & Return to Operation for 18-month Cycle ML20133F0291985-07-30030 July 1985 Safety Evaluation Accepting Util 831105 & 850605 Responses to Generic Ltr 83-28,Item 1.1 Re post-trip Review (Program Description & Procedure) ML20126F4561985-05-31031 May 1985 Safety Evaluation Supporting Util Response to Generic Ltr 83-28,Items 3.1.1,3.1.2,3.2.1,3.2.2 & 4.5.1 Re post-maint Testing Verification ML20062B8351982-07-28028 July 1982 Safety Evaluation Supporting Plant Compliance W/Esf Reset Controls Per NRC Criteria ML20126C3461980-03-20020 March 1980 Safety Evaluation Supporting Amend 51 to License DPR-30 ML20235D0971966-12-30030 December 1966 Safety Evaluation Supporting Util 660531 Proposal to Const & Operate Single Cycle BWR of 2,255 Mwt 1999-09-21
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217A9931999-09-30030 September 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data SVP-99-204, Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20217A1691999-09-22022 September 1999 Part 21 Rept Re Engine Sys,Inc Controllers,Manufactured Between Dec 1997 & May 1999,that May Have Questionable Soldering Workmanship.Caused by Inadequate Personnel Training.Sent Rept to All Nuclear Customers ML20212J0501999-09-21021 September 1999 Safety Evaluation Re Licensee Implementation Program to Resolve USI A-46 at Plant,Per GL 87-02,Suppl 1 SVP-99-179, Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20210L8661999-08-0202 August 1999 Safety Evaluation Accepting License 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs SVP-99-155, Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-99-148, Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20196H8621999-06-30030 June 1999 NRC Regulatory Assessment & Oversight Pilot Program, Performance Indicator Data, June 1999 Rept ML20195K1481999-06-16016 June 1999 Safety Evaluation Authorizing Relief Request RV-23A for Duration of Current 10 Yr IST Interval on Basis That Compliance with Code Requirements Would Result in Hardship Without Compensating Increase in Level of Quality & Safety SVP-99-123, Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20195B2591999-05-19019 May 1999 Rev 66a to CE-1-A,consisting of Proposed Changes to QAP for Dnps,Qcs,Znps,Lcs,Byron & Braidwood Stations SVP-99-104, Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-99-102, Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with1999-04-30030 April 1999 Summary Rept of Changes,Tests & Experiments Completed, Covering Period 990201-0430. with ML20205Q5291999-04-16016 April 1999 SER Concluding That Quad Cities Nuclear Power Station,Unit 1,can Be Safely Operated for Next Fuel Cycle with Weld O2BS-F4 in Current Condition Because Structural Integrity of Weld Will Be Maintained ML20205J6011999-04-0707 April 1999 Safety Evaluation Accepting Proposed Merger of Calenergy Co, Inc & Midamerican Holdings Co for Quad Cities Nuclear Power Station,Units 1 & 2 SVP-99-071, Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With1999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20205C5671999-03-19019 March 1999 Simulator Four-Yr Certification Rept ML20207D2341999-03-0101 March 1999 Post Outage (90 Day) Summary Rept, for ISI Exams & Repair/Replacement Activities Conducted 981207-1205 ML20204B1571999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Quad Cities,Units 1 & 2.With SVP-99-021, Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With1999-01-31031 January 1999 Quarterly Summary SER of Changes,Tests & Experiments Completed, Covering Period of 981101-990131,IAW 10CFR50.59 & 10CFR50.71(e).With ML20205D1311998-12-31031 December 1998 1998 Decommissioning Funding Status Rept for Yr Ending 981231 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20205M7061998-12-31031 December 1998 Unicom Corp 1998 Summary Annual Rept. with SVP-99-007, Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With1998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Quad Cities Nuclear Power Station,Units 1 & 2,IAW GL 97-02 & TS 6.9.With ML20196C8391998-11-30030 November 1998 Rev 0 to GE-NE-B13-01980-030-2, Assessment of Crack Growth Rates Applicable to Induction Heating Stress Improvement (IHSI) Recirculation Piping in Quad Cities Unit 1 SVP-98-364, Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20196G1241998-11-30030 November 1998 COLR for Quad Cities Unit 1 Cycle 16 ML20196D9651998-11-30030 November 1998 Safety Evaluation Supporting Relief Requests CR-21 & CR-24, Respectively.Relief Request CR-23,proposed Alternative May Be Authorized,Per 10CFR50.55a & Relief Request CR-22 Was Withdrawn by Licensee ML20196C8731998-11-30030 November 1998 Rev 0 to GE-NE-B13-01980-30-1, Fracture Mechanics Evaluation on Observed Indications at Two Welds in Recirculation Piping of Quad Cities,Unit 1 Station ML20196A9761998-11-20020 November 1998 Safety Evaluation Re Licensee 180-day Response to GL 95-07, Thermal Binding of Safety-Related Power-operated Gate Valves ML20196A4191998-11-19019 November 1998 Safety Evaluation Accepting QA TR CE-1-A,Rev 66 Re Changes in Independent & Onsite Review Organization by Creating NSRB SVP-98-346, Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With SVP-98-358, Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With1998-10-31031 October 1998 Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period on 980716-1031.With SVP-98-326, Monthly Operating Repts for Sept 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With1998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Quad Cities Nuclear Power Station,Units 1 & 2.With ML20153D0191998-09-18018 September 1998 Part 21 Rept Re Defect in Gap Conductance Analyses for co- Resident BWR Fuel.Initially Reported on 980917.Corrective Analyses Performed Demonstrating That Current Operating Limits Bounding from BOC to Cycle Exposure of 8 Gwd/Mtu ML20153C6771998-09-17017 September 1998 Part 21 Rept Re Defect Relative to MCPR Operating Limits as Impacted by Gap Conductance of co-resident BWR Fuel at Facilities.Operating Limit for LaSalle Unit 2 & Quad Cities Unit 2 Will Be Revised as Listed ML20151T2711998-09-0404 September 1998 Safety Evaluation Accepting Licensee Response to NRC Bulletin 95-002 ML20151Y7261998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for Quad Cities Nuclear Power Station ML20237E2331998-08-21021 August 1998 Revised Pages of Section 20 of Rev 66 to CE-1-A, QA Topical Rept ML20151Y7301998-07-31031 July 1998 Revised MOR for Jul 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20237A6251998-07-31031 July 1998 Monthly Operating Repts for July 1998 for Quad Cities Nuclear Power Station,Unit 1 & 2 SVP-98-328, Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period of 971001-980715,per 10CFR50.59 & 10CFR50.71(e).With1998-07-15015 July 1998 Summary Rept of Changes,Tests & Experiments Completed, Including SEs Covering Period of 971001-980715,per 10CFR50.59 & 10CFR50.71(e).With SVP-98-249, Monthly Operating Repts for June 1998 for Quad Cities Nuclear Power Station,Units 1 & 21998-06-30030 June 1998 Monthly Operating Repts for June 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 SVP-98-215, Monthly Operating Repts for May 1998 for Quad Cities Nuclear Power Station Units 1 & 21998-05-31031 May 1998 Monthly Operating Repts for May 1998 for Quad Cities Nuclear Power Station Units 1 & 2 ML20247N6281998-05-19019 May 1998 Rev 2 to COLR for Quad Cities Unit 2 Cycle 15 ML20216C0561998-04-30030 April 1998 Safe Shutdown Rept for Quad Cities Station,Units 1 & 2, Vols 1 & 2.W/22 Oversize Figures SVP-98-176, Monthly Operating Repts for Apr 1998 for Quad Cities Nuclear Power Station,Units 1 & 21998-04-30030 April 1998 Monthly Operating Repts for Apr 1998 for Quad Cities Nuclear Power Station,Units 1 & 2 ML20217D0281998-04-22022 April 1998 Part 21 Rept Re Additive Constants Used in MCPR Determination for Siemens ATRIUM-9B Fuel by Core Monitoring Sys Were Found to Be non-conservative.SPC Personnel Notified All Customers w/ATRIUM-9B Lead Test Assemblies ML20217G3951998-04-0808 April 1998 TS 3/4.8.F Snubber Functional Testing Scope Quad Cities Unit 2 TS (Safety-Related) Snubber Population 129 Snubbers SVP-98-128, Monthly Operating Repts for Mar 1998 for Quad Cities Nuclear Station Units 1 & 21998-03-31031 March 1998 Monthly Operating Repts for Mar 1998 for Quad Cities Nuclear Station Units 1 & 2 1999-09-30
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j Enclosure 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PLANT SYSTEMS BRANCH INTERIM COMPENSATORY MEASURES AND REQUEST FOR EXEMPTION FROM 10 CFR PART 50, APPENDIX R, SECTION III.G REQUIREMENT
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REGARDING HOT SHUTDOWN REPAIRS FOR A FIRE EVENT IN THE PLANT QUAD CITIES, UNITS 1 AND 2 DOCKET NOS. 50-254 and 50-265
1.0 INTRODUCTION
On December 30, 1982, the NRC issued a Safety Evaluation Report (SER) relating to Sections III.G.3 and III.L of 10 CFR Part 50, Appendix R (alternative / dedicated shutdown capability for a reactor following a t ire event in the plant) for ,
Quad Cities, Units 1 and 2, wherein the staff concluded that the plant-net the requirements of the above sections with regard to alternative shutdown capability.
Subsequently, by letter dated December 18, 1984, Commonwealth Edison, the licensee for the plant, submitted an Appendix R reevaluation report stating that it was necessitated by Generic Letter 83-33, dated October 19, 1983 which defined NRC staff positions on certain Appendix R requirements. In the above submittal, the licensee identified the Interiin Compensatory Measures (ICMs) needed to ensure safe shutdown of the plant following a fire event in the plant during the interim period (i.e., until the permanent hardware modifications are completed). The report additionally contained a request-for exemption from specific III.G requirements relating to fire protection features for select areas. Based on a review of the submittal, the staff has determined that the safe shutdown capabilty including the alternative shutdown capability at the plant continues to be essentially the same as that described by the licensee in their earlier submittals. The staff has, therefore, determined that its earlier acceptance (December 30, 1982 SER) remains valid.
The staff, however, sought information relating to fire-induced high impedance faults and electrical isolation deficiency concerns which can compromise safe shutdown capability, since these were not explicitly addressed in the reevaluation. The staff also requested additional information on the ICMs required to ensure safe shutdown capability in the interim period. By letters dated December 30, 1986 January 12,1987, March 13,1987, July 15,1987, September 30, 1987, October 1,1987, October 9,1987, and November 20, 1987, the licensee provided their responses. In the these submittals, the licensee proposed some manual operations including hot shutdown repairs to eliminate fire-induced electrical isolation deficiencies, spurious operations and high impedance f aults. Also, the licensee requested exemptions from the Appendix R,Section III.G.1 requirement for performing repairs for achieving and maintaining hot shutdown, in so far as it is interpreted as disallowing such repairs. In the March 13, 1987 submittal, the licensee further stated that since all the needed safe shutdown hardware modifications had been completed, their corresponding ICMs would not be needed. Also, by the July 17, 1987 submittal, the licasee identified a few differences relating to the plant safe shutdown configuration as it exists now from what has been described in the earlier E (December 30,1982). For the reasons stated above, this SER addresses only differences from the earlier SER and the licensee's reevaluation relating to fire-induced electrical deficiency concerns, spurious operations concerns and high impedance faults concerns. Another SER, to be provided at a later date, will address technical exemptions requested in the reevaluation report related to fire protection features for specific plant areas.
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4 2.0 EVALUATION' 2.1 Spurious Operations and High Impedance Faults ,
In their submittals, the licensee stated that a fire in any one of certain plant areas could damage RHR system logic cables associated with safe shutdown equipment which, in turn, could result in spurious operations of RHR pumps and valves, diesel generators auxiliary equipment, safety relief valves (SRVs) and 4 kv breakers. Additionally, a fire event in 'any one of certain plant areas could damage the circuits for SRVs resulting in their spurious operations. To eliminate these spurious operations, the licensee has proposed to deenergize applicable circuits by opening respective breakers at de distribution panels located in Fire Areas (FA) TB-I and TB-III(TurbineBuildingNorthernandSouthernZoneGroups). For a fire, i in either FA TB-1 or TB-III, the licensee has proposed to deenergize these circuits by pulling out control power fuses located in the applicable two of four panels in a timely manner (8 fuses within 30 minutes after scram for handling the RHR logic circuit concern and 10 fuses within 10 minutes after scram for handling the SRVs concern). All four panels, of anels (one for each unit) contain 8 fuses each and the which twopanels other two of the p(one for each unit) contain 10 fuses each, are located outside FAs TB-I and TB-III and are easily accessible following a fire event in either FA TB-I or TB-III.
Regarding fire-induced high impedance faults (faults in circuits supplying power to non-safe shutdown loads from a common power source that supplies power also to safe shutdown loads) which can affect power '
supply to safe shutdown loads, the licensee stated that plant safe shutdown procedures require the operator to shed all non-safe shutdown loads from common power buses by tripping manually the associated breakers in a timely manner. Additionally, these procedures require pulling out the 124 V de control power fuses for electrically operated breakers associated with non-safe shutdown loads that are supplied power by 480 V or 4 kv switchgear common buses. This task will be perforn'ed prior to tripping applicable breakers as a precaution against their possible spurious closures. The licensee pointed out that such fuse pulling would be perforried either within 30 minutes or 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after scram depending upon whether such actions are required before initiating reactor water makeup (30 minutes) or suppression pool cooling (3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />).
With regard to the fuse pulling operations mentioned above, the licensee stated that applicable control power fuses are easily identifiable, i
readily accessible, easy to remove, under periodic surveillance, and that their removal would not involve any significant operator hazard. The licensee further stated that the plant shutdown procedures include operator instructions to perform the above tasks in a timely manner.
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Based on the above, the staff finds the licensee's proposed manual actions, i.e., tripping the applicable breakers and pulling out the applicable fuses in a timely manner for handling spurious operation and high impedance fault concerns, to be acceptable. The staff further ~
recommends that the licensee's request for exemptions from the Appendix R,Section III.G.1 requirement for performing the above mentioned hot shutdown repair, i.e., fuse pulling for achieving and maintaining hot shutdown, be granted.
2.2 Electrical Isolation Deficiency Regarding the fire-induced electrical isolation deficiency (i.e., a fault on a remote circuit blowing a fuse common to both local and remote control circuits, prior to isolation of the needed hot shutdown circuit), that can compromise the ability to transfer the needed hot shutdown circuit to local control, the licensee has identified seven cables as vulnerable to this design problem. This is because these cables which are part of the 125V de control circuitry for four breakers at the 480V buses, are singely fused. The licensee stated that, in the event the common control power fuses associated with these breakers are fire damaged and additionally these breakers are found open (two of these breakers are normally closed and may not require any manual action), the plant shutdown procedures will require them to be manually closed in a timely manner (30 minutes) using a jacking' handle located in a cabinet in the vicinity of the applicable 480V'switchgears. The licensee further stated that the maximum number of breakers that may require such manual closing at any one time due to a fire event is three. Besides the above, the licensee ,
has identified three other control circuits, associated with engine starting controls for the Unit 1. Unit 2 and swing diesel generators, as vulnerable to electrical isolation deficiencies. The licensee stated that, for these circuits, all applicable blown fuses would be replaced in a timely manner (within 30 minutes) and no more than two blown fuses, at any one time, would require such replacement. The licensee has committed to maintain replacement fuses and fuse pullers under surveillance in proximity of the engine starting controls for the diesel generators, .
and provide emergency lighting and needed man power in these areas to '
facilitate fuse replacements in a timely manner. The licensee further claimed that the circuits involved are low voltage control circuits and the fuses, though rated at 15 amperes, will actually carry much less current. Therefore fuse replacement will not pose any undue operator hazard. Based on the above, the staff has determined the licensee's proposed manual closing of applicable breakers and hot shutdown repairs, (i.e., fuse replacement) meet the intent and purpose of IE Information Notice No. 85-09, " Isolation Transfer Switches and Post-Fire Shutdown a Capability", dated January 31, 1985, and are, therefore, acceptable. The staff further recommends that the licensee's request for exemption from ;
Appendix R,Section III.G.1 requirement for performing aforementioned hot )
shutdown repairs (i.e., fuse replacement) for acheiving and maintaining j hot shutdown, be granted. j l
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.2.3 Differences with December 30, 1982 SER In the July 17, 1987 submittal, the licensee identified the differences '
in the safe shutdown configuration as it exists now at the plant from ~
what has been described in the earlier SER dated December 30, 1982.
} The licensee additionally provided supporting , justification for these differences in the above submittal end other submittals referred to in this SER. These differences are listed below:
- 1. Backup water supply source for the safe shutdown makeup pump will be provided by the Fire Water System (FWS) instead of the Service Water System as originally indicated in the earlier SER Section 3.1.2.
Based on their hydraulic evaluation on the adequacy of the FWS, the licensee has concluded that the system can simultaneously meet the maximum fire demand and supply cooling water to the safe shutdown makeup pump room cooler, and also provide backup water supply source for the safe shutdown makeup pump at later times when needed.
- 2. RHR flow indication instrumentation included as being available during a fire event, in Section 3.1.5 of the earlier SER, is not considered as necessary diagnostic instrumentation. However, during torus cooling, the needed diagnostic instrumentation will be provided by suppression pool temperature indication and RHR pump discharge pressure indication.
- 3. Earlier SER Section 3.3 indicated there will be no need for hot or cold shutdown repairs for achieving and maintaining safe shutdown.
However, as indicated in Sections 2.1 and 2.2 of this SER, there may be hot shutdown repairs (i.e., fuse pulling and/or fuse replacement) depending upon the fire event. Cold shutdown repairs may also be needed for certain fire events (these are described in Section 2.4 of the licensee's December 18, 1984 submittal).
- 4. The plant does not have documentation for breaker / fuse coordination for all instrumentation and power circuits as implied in the earlier SFR Section 3.4.1. However, plant safe shutdown procedures include '
operator instructions for shedding non-safe shutdown loads from common power sources, and for fuse pulling when needed to handle high impedance faults associated with certain common power sources. ,
These insure all the safe shutdown loads in a given bus are free of l fire induced faults whenever the bus is utilized to power safe shutdown loads.
Based on the above, the staff has determined there is reasonable assurance these differences will not compromise the safe shutdown capability of the plant and are, therefore, acceptable.
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3.0 CONCLUSION
LThe staff coycludes that the licensee's proposed approaches for resolving' ~
the fire-induced. concerns (i.e.,- spurious operations identified in this SER,' high impedance faults, and electrical isolation' deficiency) are acceptable. Consequently the staff-recommends tFat the licensee's exemption-requests to. allow conducting aforementioned hot shutdown repairs (i.e., fuse pulling and/or fuse replacement), for achieving and maintaining hot shutdown, be granted. . Futhermore,t theistaff concludes that-the differences between the present safe shutdown configuration at the plant'from what'has been described
.in.the December 30, 1982 SER, with regard to'those items listed in Sectie. 2.3 of this SER. are acceptable.
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