|
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217G2161999-10-15015 October 1999 Errata Pages 2 & 3 for Safety Evaluation Supporting Amend 168 Issued to FOL DPR-63 Issued on 990921.New Pages Change Description of Flow Control Trip Ref Cards to Be Consistent with Application for Amend ML20207G2261999-06-0707 June 1999 SER Accepting Proposed Mod to Each of Four Core Shroud Stabilizers for Implementation During Current 1999 Refueling Outage at Plant,Unit 1 ML20206U5351999-05-17017 May 1999 SER Accepting GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Plant, Units 1 & 2 ML20155E2001998-11-0202 November 1998 Safety Evaluation Approving NMP 980227 Request for Extension of Reinspection Interval for Core Shroud Vertical Welds at NMP1 from 10,600 Hours to 14,500 Hours of Hot Operation ML20154D8401998-10-0505 October 1998 Safety Evaluation Accepting Proposed Changes Related to PT Limits in Plant,Unit 1 TSs ML20217F4341998-03-19019 March 1998 SER Related to Proposed Restructuring New York State Electric & Gas Corp,Nine Mile Point Nuclear Station,Unit 2 ML20198H9941997-12-29029 December 1997 SE Supporting Approval of Application Re Long Island Power Authority Aquisition of Long Island Lighting Co,Subject to Discussed Condition ML20197C4771997-01-0303 January 1997 Safety Evaluation Supporting Nine Mile Point Unit 1 Reactor & Turbine Building Blowout Panels ML20197C4541997-01-0202 January 1997 Safety Evaluation Supporting Resolution of Nine Mile Point Reactor/Turbine Building Pressure Relief Panel Outside Design Basis Issue ML20056H4751993-08-27027 August 1993 Safety Evaluation Accepting Licensee Proposal for Continued Insp & Repair of Flaw in Weld Joining HPCS Nozzle Safe to safe-end Extension ML20247K2531989-09-11011 September 1989 Safety Evaluation Supporting Amends 123 & 41 to Licenses DPR-61 & NPF-49,respectively ML20247E3761989-09-0707 September 1989 Safety Evaluation Supporting Amends 122,34,143 & 40 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20248C0731989-08-0303 August 1989 Sser Accepting 880601,0909 & 890602 Changes to ATWS Mitigation Sys Actuation Circuitry for Plants ML20246L2571989-06-26026 June 1989 Safety Evaluation Supporting Amends 118,33,142 & 36 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20245J0751989-04-25025 April 1989 Safety Evaluation Supporting Amends 114,30,141 & 33 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20205R1311988-10-31031 October 1988 Safety Evaluation Supporting Amend 101 to License DPR-63 ML20205Q5721988-10-31031 October 1988 SER Re Const Mod & Licensing of Company Facility-1 Cpdf. Licensee Technically Qualified to Modify Existing Facility in Such Way as to Assure Adequate Protection of Common Defense & Security ML20205Q5761988-10-31031 October 1988 SER Re Application for CP for Alchemie Facility-2 Oliver Springs.Licensee Technically Qualified to Construct & Operate Proposed Facility in Such Way as to Assure Adequate Protection for Common Defense & Security ML20205M5731988-10-26026 October 1988 Safety Evaluation Supporting Amends 108,25,134 & 26 to Licenses DPR-61,DPR-21,DPR-65 & NPF-49,respectively ML20155G4801988-09-28028 September 1988 Safety Evaluation Supporting Amends 107,23,132 & 24 to Licenses DPR-61,DPR-21,DPR-65 & NPF-24,respectively ML20195G5531988-06-24024 June 1988 Safety Evaluation Accepting Proposed Reracking of Util Spent Fuel Storage Pools from Criticality Standpoint.Enrichment of Fuel to 4.5 Weight % U-235 May Be in Conflict W/ 10CFR51 Table S4 & Should Be Investigated by NRC ML20154H2171988-05-18018 May 1988 Safety Evaluation Accepting Util 880414 Submittal Re Reload Startup Physics Test Program ML20147D9631988-02-25025 February 1988 Safety Evaluation Accepting Util 860404 Evaluation of Environ Qualification of Equipment Considering Superheat Effects of high-energy Line Breaks for Plants,Per IE Info Notice 84-90 ML20147E9261988-02-23023 February 1988 Safety Evaluation Supporting Amends 100,14,125 & 15 to Licenses DPR-61,DPR-21,DPR-65 & NPF-45,respectively ML20236M9631987-11-0606 November 1987 Safety Evaluation Accepting Util Proposed ATWS Mitigating Sys Actuation Circuitry for Facilities,Per 10CFR50.62(c)(1) & Pending Final Resolution of Tech Spec Issue ML20236L2001987-10-30030 October 1987 Safety Evaluation Supporting Amends 11 to Licenses NPF-37 & NPF-66,respectively & Amend 1 to License NPF-72 ML20195G5501987-10-28028 October 1987 Safety Evaluation Accepting Proposed Reracking of Util Spent Fuel Storage Pool.Licensee May Be Required to Perform Addl Analyses to Verify Maint of Required Subcriticality If Staff Boraflex Studies Reveal Gap Development in Subj Boraflex ML20235A7331987-09-18018 September 1987 Safety Evaluation Re Installation of Alternate Rod Injection (ARI) Sys & Adequacy of Plant Reactor Coolant Recirculating Pump Trip (RPT) Sys,In Compliance W/Atws Rule 10CFR50.62. ARI & RPT Acceptable ML20195G5461987-08-21021 August 1987 Safety Evaluation Accepting Proposed Util Tech Specs Change to Increase Max Allowed Enrichment of Reload Fuel to 4.2 Weight % U-235 ML20236Y0221987-07-0808 July 1987 Safety Evaluation Clarifying Determination of Acceptability of Test Duration for Performance of Integrated Leak Rate Test at Plant ML20214M3361987-05-22022 May 1987 Safety Evaluation Supporting Util Rept Entitled, Rod Swap Methodology Rept for Startup Physics Testing ML20214M1491987-05-21021 May 1987 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 2.1 (Part 2) Re Established Interface W/Nsss or Vendors of Each Component of Reactor Trip Sys ML20153F3851987-05-0404 May 1987 SER Accepting Util 870301 & 0407 Requests for Rev to Tech Specs,Modifying MSIV Actuation Control Sys,Per IEEE Std 279 & GDC 21 ML20153F3671987-04-30030 April 1987 Safety Evaluation Accepting Util 870311 Proposed Changes to Tech Spec Tables 2.2.1-1,3.6.1.2-1 & 3.6.3-1 Re MSIV-closure Setpoint ML20238C7121987-04-0808 April 1987 Safety Evaluation Concluding That Facility May Resume Operation W/O Leakage Control Sys But w/post-accident Leakage Mgt ML20207R6481987-03-11011 March 1987 Safety Evaluation Supporting Licensee Response to Generic Ltr 83-28,Item 2.1 (Part 1) Equipment Classification.Program Acceptable ML20212J4041987-03-0202 March 1987 Safety Evaluation Re Util 831104 Response to Generic Ltr 83-28,Item 4.5.2, Reactor Trip Sys Reliability,On-Line Testing. Licensee Meets NRC Position ML20212B5671987-02-24024 February 1987 Safety Evaluation Re Util Response to Generic Ltr 83-28,Item 4.5.2, Reactor Trip Reliability On-Line Testing. Response Acceptable ML20211Q2971987-02-18018 February 1987 Safety Evaluation Re Auxiliary Feedwater Sys Reliability (Generic Issue 124) for Prairie Island Units 1 & 2 ML20211G4361987-02-14014 February 1987 Safety Evaluation Supporting Util 831110 Response to Generic Ltr 83-28,Item 4.5.2 Re on-line Testing for Reactor Trip Sys Reliability ML20153F3511987-02-0707 February 1987 Safety Evaluation Accepting Util Latest Design Mods to MSIV Actuation Control Sys,Per IEEE 279 & GDC 21 ML20210R2061987-02-0606 February 1987 Safety Evaluation Supporting Util 850517,0802,0823,1211 & 860429 Submittals Re Environ Effects of High Energy Line Breaks in Auxiliary Steam or Steam Generator Blowdown Sys. Design of Blowdown Sys Acceptable ML20210T2571987-02-0606 February 1987 SER Re Util 850802 Submittal Describing Design Details of Steam Generator Blowdown & Auxiliary Steam Sys to Detect & Isolate High Energy Line Breaks.Sys Design Acceptable, However,Two Deviations from IEEE-STD-297 Criteria Apparent ML20238C6501987-02-0505 February 1987 Safety Evaluation Accepting MSIVs for Facility Operation Up to First Refueling,Contingent on Successful Completion of Preoperational Tests & Prototype Testing Program.Briefing for Commissioner Asselstine Re MSIVs & SALP Input Also Encl ML20238C5801987-02-0202 February 1987 Safety Evaluation Concluding That Facility May Resume Startup Testing & Operation for First Operating Cycle W/ Installed Refurbished Msivs.Necessity of Addl Leak Testing to Be Determined After Review of Prototype Test Results ML20209B1511987-01-28028 January 1987 SER Supporting Util Responses to Generic Ltr 83-28,Item 4.5.2 Re Reactor Trip Sys on-line Testing ML20209C3571987-01-23023 January 1987 SER Supporting Facility Design,Per Generic Ltr 83-28,Item 4.5.2, Reactor Trip Sys Reliability,On-Line Testing ML20209C2151987-01-21021 January 1987 Safety Evaluation Re Auxiliary Feedwater Sys Reliability (Generic Issue 124) at Prairie Island Units 1 & 2.Util Actively Pursuing Improvements in Sys Reliability & Reducing Sys Challenges ML20215E0061986-12-12012 December 1986 Safety Evaluation Supporting Util Responses to Generic Ltr 83-28,Item 2.1 (Part 1) Re Equipment Classification of Reactor Trip Sys Components ML20214J9911986-11-24024 November 1986 SER Re Conformance to Generic Ltr 83-28,Item 2.1 (Part 1) on Equipment Classification (Reactor Trip Sys Components) at Selected GE BWR Plants.Part 1 Requirements Met 1999-06-07
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G2161999-10-15015 October 1999 Errata Pages 2 & 3 for Safety Evaluation Supporting Amend 168 Issued to FOL DPR-63 Issued on 990921.New Pages Change Description of Flow Control Trip Ref Cards to Be Consistent with Application for Amend ML20216J9251999-09-30030 September 1999 Suppl to Special Rept:On 990621,11 Containment Hydrogen Monitoring Sys Chart Recorder Was Indicating Below Normal Operating Range.Caused by Excessive Wear on Valve Body & Discs of Bypass Pump.Sample Pump Replaced ML20217K4631999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Nine Mile Point, Unit 1.With ML20212F7301999-09-21021 September 1999 Special Rept:On 990907,CR Operators Declared 12 Containment Hydrogen Monitoring Sys Inoperable for Planned Maint.Cause of Low Flow Condition Was Determined to Be Foreign Matl. Replaced Sample Pump Valve Discs ML20212B9081999-09-14014 September 1999 Special Rept:On 990901, 12 Containment Hydrogen Monitoring Sys Was Declared Inoperable for Planned Maint.Caused by Planned Maint Being Performed as Corrective Action.Check Valves with O Rings Were Replaced ML20216F5141999-08-31031 August 1999 Rept on Status of Public Petitions Under 10CFR2.206 ML20212C4601999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Nine Mile Point Nuclear Station,Unit 1.With ML20210U4591999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Nine Mile Point, Unit 1.With ML20209D0291999-07-0202 July 1999 Special Rept:On 990621,operator Identified That Number 11 Hydrogen Monitoring Sys (Hms) Chart Recorder Was Indicating Below Normal Operating Range.Cause Indeterminate.Licensee Will Complete Troubleshooting of Subject Hms by 990709 ML20210B9081999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Nine Mile Point Unit 1.With ML20209F8811999-06-0808 June 1999 Rev 1 to NMP Unit 1 COLR for Cycle 14 ML20207G2261999-06-0707 June 1999 SER Accepting Proposed Mod to Each of Four Core Shroud Stabilizers for Implementation During Current 1999 Refueling Outage at Plant,Unit 1 ML20196E2111999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Nmp,Unit 1.With ML20207B0241999-05-18018 May 1999 Safety Evaluation of Topical Rept TR-107285, BWR Vessel & Intervals Project,Bwr Top Guide Insp & Flaw Evaluation Guidelines (BWRVIP-26), Dtd December 1996.Rept Acceptable ML20206U5351999-05-17017 May 1999 SER Accepting GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Plant, Units 1 & 2 ML20196L2301999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Nmp,Unit 1.With ML20205L0541999-04-0101 April 1999 Nonproprietary Replacement Pages to HI-91738,consisting of Section 5.0, Thermal-Hydraulic Analysis ML20205S5701999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for NMP Unit 1.With ML20207M9231999-03-12012 March 1999 Amended Part 21 Rept Re Cooper-Bessemer Ksv EDG Power Piston Failure.Total of 198 or More Pistons Have Been Measured at Seven Different Sites.All Potentially Defective Pistons Have Been Removed from Svc Based on Encl Results ML20207G2671999-03-0101 March 1999 Special Rept:On 990315,Nine Mile Point,Unit 1 Declared Number 12 Containment Hydrogen Monitoring Sys Inoperable. Caused by Degraded Encapsulated Reed Switch within Flow Switch FS-201.2-1495.Technicians Replaced Flow Switch ML20204C9971999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Nine Mile Point,Unit 1.With ML20207E9311999-02-26026 February 1999 Part 21 Rept Re Sprague Model TE1302 Aluminum Electrolytic Capacitors with Date Code of 9322H.Caused by Aluminum Electrolytic Capacitors.Affected Capacitors Replaced ML17059C5501999-01-31031 January 1999 Rev 0 to MPR-1966(NP), NMP Unit 1 Core Shroud Vertical Weld Repair Design Rept. ML20199M0891999-01-22022 January 1999 Part 21 Rept Re Failure of Square Root Converters.Caused by Failed Aluminum Electrolytic Capacitory Spargue Electric Co (Model Number TE1302 with Mfg Date Code 9322H).Sent Square Root Converters Back to Mfg,Barker Microfarads,Inc ML20199K9331998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Nine Mile Point Nuclear Station,Unit 1.With ML20210R8441998-12-31031 December 1998 1998 Annual Rept for Energy East ML20206P2391998-12-31031 December 1998 Special Rept:On 981222,operators Removed non-TS Channel 12 Drywell Pressure Recorder & Associated TS Pressure Indicator from Svc.Caused by Intermittent Measuring Cable Connection in non-TS Recorder Circuitry.Replaced Cable ML20206P2421998-12-30030 December 1998 Special Rept:On 981219,number 12 Hydrogen Monitoring Sys (Hms) Was Declared Inoperable When Operators Closed Valve 201.2-601.Caused by Indeterminate Failure of Valve 201.2-71. Supplemental Rept Will Be Submitted After Valve Is Repaired ML20198M3571998-12-23023 December 1998 Special Rept:On 981210,operators Declared Number 11 Inoperable,Due to Failure of CR Chart Recorder.Caused by Inverter Board in Power Supply Circuitry of Recorder Due to Component Aging.Maint Personnel Replaced Failed Inverter ML20198D9361998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Nine Mile Point,Unit 1.With ML20155E2001998-11-0202 November 1998 Safety Evaluation Approving NMP 980227 Request for Extension of Reinspection Interval for Core Shroud Vertical Welds at NMP1 from 10,600 Hours to 14,500 Hours of Hot Operation ML20195J4141998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Nine Mile Point Nuclear Station,Unit 1.With ML20154D8401998-10-0505 October 1998 Safety Evaluation Accepting Proposed Changes Related to PT Limits in Plant,Unit 1 TSs ML20154P1821998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Nine Mile Point Nuclear Station,Unit 1.With ML20153B2001998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Nmpns,Unit 1.With ML20237C6351998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Nine Mile Point Nuclear Station,Unit 1 ML20236T5911998-07-20020 July 1998 LER 98-S01-00:on 980618,security Force Member Left Nine Mile Point,Unit 2 Vehicle Gate Unattended Without Ensuring,Gate Alarm Had Been Reactivated.Caused by Inadequate Work Practice.Vehicle Gate Alarm Was Activated ML18040A3491998-07-0202 July 1998 LER 98-017-00:on 980602,control Room Ventilation Sys Was Declared Inoperable.Caused by Original Design Deficiency. Mod Designed,Tested & Implemented Prior to Startup from RF06 to Correct Design deficiency.W/980702 Ltr ML20236Q1701998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Nine Mile Point Nuclear Station,Unit 1 ML17059C1011998-06-24024 June 1998 LER 98-014-00:on 980525,noted Differences Between Actual Valve Weights & Weights Shown on Engineering Drawings.Caused by Vendor Failing to Provide Accurate Valve Weights.Revised Valve Drawings & Associated Calculation,Per 10CFR21 ML20151P1751998-06-16016 June 1998 Rev 0 to SIR-98-067, Evaluation of NMP Unit 2 Feedwater Nozzle-to-Safe End Weld Butter Indication (Weld 2RPV-KB20, N4D) ML18040A3451998-06-0404 June 1998 LER 98-004-01:on 980302,TS Required LSFT of Level 8 Trip of Main Turbine Was Missed.Caused by Knowledge Deficiency of EHC Sys.Revised Applicable LSFT Procedures Prior to Refueling Outage 6.W/980604 Ltr ML20249B4971998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Nine Mile Point Nuclear Station,Unit 1 ML20248F3531998-05-21021 May 1998 Part 21 Rept Re Electronic Equipment Repaired or Reworked by Integrated Resources,Inc from Approx 930101-980501.Caused by 1 Capacitor in Each Unit Being Installed W/Reverse Polarity. Policy of Second Checking All Capacitors Is Being Adopted ML20198B4991998-05-15015 May 1998 Non-proprietary Replacement Pages for Attachment F to Which Proposed to Change TS 5.5, Storage of Unirradiated & Sf ML20247R1141998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Nine Mile Point Nuclear Station,Unit 1 ML20217B0621998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Nine Mile Point Nuclear Station,Unit 1 ML20217F4341998-03-19019 March 1998 SER Related to Proposed Restructuring New York State Electric & Gas Corp,Nine Mile Point Nuclear Station,Unit 2 ML17059C1681998-03-19019 March 1998 Revised Niagara Mohawk Powerchoice Settlement Document for NMPC PSC Case Numbers 94-E-0098 & 94-E-0099, Vols 1 & 2 ML17059B9051998-02-28028 February 1998 NMP Unit 1 Boat Samples Analyses Part Iii:Tension Tests, RDD:98:55863-004-000:01 1999-09-30
[Table view] |
Text
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1
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'.W f, g NUCLEAR REGULATORY COMMISSION
% l WASHINGTON, D. C. 20555
%,*****/
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO APPENDIX J TO 10 CFR 50 TESTING NINE MILE POINT UNIT 2
, DOCKET NO. 50-410 l
INTRODUCTION 1
By letter dated July 3,1986, the applicant withdrew a request for exemption l from Section !!! C of Appendix J to 10 CFR 50 for 16 relief valves. The exemption request was to eliminate the need to locally leak rate test these vdives per the requirements identified under the type C test program. The withdrawal was based on several factors. Three of the valves were determined to be capable of reverse flow testing. As a result, these valves will be type C air tested in accordance with Appendix J. The remaining 13 valves will have their discharge lines modified, prior to fuel load, so that they do not represent a conta'nment atmospheric leak path, Specifically, the vacuum breakers will be seal welded closed. This modification eliminates the pathway
- to the containment atmosphere since the discharge pipes end within the suppression pool and below the minimum post-l.oca drawdown water level.
EVALUATION The staff has reviewed the requested exemption withdrawal for 16 relief valves from Section !!! C of Appendix J to 10 CFR 50. The applicant has reevaluated the potential of reverse testing. The results have enabled ;om to include that for three valves, the reverse test is as conservative as a forward test.
Therefore, these three valve will be tested in the reverse direction, which is in' compliance with the requirements of Appendix J.
The remaining 13 valves with their associated piping will be modified, prior to fuel load, to eliminate them as potential containment atmosphere leak pathways. This will be accomplished by seal welding closed the discharge line vacuum breakers. The weld will be continuous and leak checked to assure a leak tight barrier. In addition, discussions with the applicant have indicated the elimination of the vacuum breaker function will not cause steam condens'ation loads to exceed design. Af ter these modifications have been made, the 13 relief valve can be assumed to qualify for hydrostatic rather than pneumatic testing. As a result, Appendix J requirements are not applicable. i Therefore, an exemption from the Type C testing requirements is not required.
l CONCLUSION The staff concurs with the approach taken by the applicant to withdraw the exemption request for 16 relief valves. For three valves, the reverse dirQqdQG Mih RM thMQ Vahts ain q99S199 (QPSlitnte HRh Appendh 4 nwineenu, henfon no empion u needto, foe 30 neening D MHet valves, the corm 11tted to modificat';as would make Appendix J requirements inapplicable. Therefore, the exemption request is not necessary, 8809070370 000010 PDR FOIA L KUDLICKBO-356 PDR . l
,g.
.,bf
, TABLE 3.6.3-1 (Continued)
~
M PRIMARY CONTAINMENT ISOLATION MtVES 2i ISOLATION VALVE ISOLATION MAXIMUM CLOSING S
z VALVE NO. VALVE FUNCTION GROUP SIGNAL (a) TIME (SECONDS)
) D. Other '
$ Safety Relief 2Ril5*RV20 2RilS*RV61 ,
[A,t(s).(d)
,C(o).(d)
Ril5Ril5 RvRv disch.
disch. to SP to SP Outside Outside IVsIVs 2RilS*RV108(el (d)< Ril5 Rv disch. to SP Outside IVs 2Ril5*RV110(o). (4):e SDC to RilR Pump suction Rv 2RilS*RV139(o) (d) s RHR lidr. Flush to Radwaste RV ~
2RilS*RV152(r) (c) SDC Supply from RCS RV Inside IV 2RilS*RV56 A B(d) RHS fiX shell side RVs ,
y 2RHS*5V34 A,B(d) Riis HX steam supply Safety valves -
Ril5 HX steam supply Safety valves 2 2RilS*5V62 A B(d) 2RilS*RVV35 A,B(d) RHS Vacuum Breakers S 2CSL*RV105(o). (d) < CSL RV Disch. to SP Outside IV 2CSL*RV123(o) (4) CSL RV Disch. to SP Gutside IV 2RilS*RVV36 A,B(d) RilS Vacuum Breakers 2CCP*RV170(o).(n) CCP RV Discharge Inside IV q ex --a 2CCP*RV171(e) (n) CCP RV Discharge Inside IV Z
2C51t* RVll3(cl (d) C511 RV Disch. to SP Outside IV '
y 2C51L*RV114(o). (d) C5il RV Disch. to SP Outside IV D
E ===l
=-
ce e$A
. m-n n 39g;p.
=,
TABLE 3.6.3-1 (Continued) h m
@ lMARY CONTAINMENT ISOLATION VALVES 1
? ISOLATI05 VALVE ISOLATION MAXIMUM CLOSING
[ VALVE NO. VALVE FUNCTIOfDN GROUP' SIGNAL (a) TIME (SECONOS) o
$ Check Valves ,
- 2RHS*A0 VIE A,B,C(h) RHS/LPCI to RP4@V Inside IVs g 2RHS*A0V39 A,B(h) l SDC to RCS Inghlde IVs
" 2 CPS *V5s Nitrogen SupPjMy to 2 CPS *A0VIO7 Inside IV 2 CPS *V51 Mitrogen SupP),My to 2 CPS *A0VIO9 Inside IV 2C5H*A0W108(h) CSH to RPV Ingkide IV 2CSL*A0W181(h) CSL to RPV Inikide IV 21CS*A0V156(h) ICS to RPV OutJt' side IV
} 2ICS*A0V157(h) ICS to RPV In% hide IV I 2SLS*Vis SLS to RPV Ingkide IV U '
2GSN"V17e N2 Purge to Tif1p Index Mech. Inside IV 2!AS*V443 IAS to ADS AcfMunulators Inside IV 2IAS*V4c3 IAS to ADS AcfMtsoulators Inside IV 2RCS*V59 A B RDS to RCS Punikp A Seal Outside IVs 2RCS*f68 A,B RDS to RC5 Pu%p A Seal Inside.IVs q 2RCS*V98 A,8 RDS to RCS Pu%p A Seal Outside IVs %
y~-
2RHS*VIS(d)(f) Discharge ChefAk from RCIC to Supp. Pool b 2RHS*V29(d)(f) Discharge ChefAk from RCIC to Supp. Poo'. y 2RHS*V117(d)(f) Check Valve f9Yom RCIC Drain to Supp. Pool 2RHS*V118(d)(f) Check Valve f4Yom RCIC Drain to Supp. Pool g 2FWS*A0v23 A,B(h) feedwater toffkPV Outside IV's U
2FWS*VI2 A,B Feedwater toflRPV Inside IV's h -
ee ~
tJe
- p - f Q $..uf . o .
~
.l
. TABLE 3.6.3-1 (Continued) k PRIMARY CONTAINND.'T ISOLATION VALVES 5
ISOLATION VALVE ISOLATION MAXIMUM Ct.051NG S VALVE NO. VALVE FUNCTION GROUP SIGNAL (a) TIME.(SECONDS)
"i
, Excess Flow Check (e) c Reactoi Instrumenta g
-e tion Lines
- 215C"EFV1 Inst. Line from MSS 215C*EFV2 Inst. Line from N14,200* .
215C*EFV3 Inst. Line from N14,160*
,- 215C*EFV4 Inst. Line from M13,190*
. 2ISC"EFV5 Inst. Line from N14,20' R 2ISC*EFV6 Inst. Line from N14,340*
- - 2ISC*EFV7 Inst. Line from N13,10' T 2ISC"EFV8 Inst. Line from M12,160*
l
- g 2ISC*EFV?D Inst. Line from N12.200*
2ISC*EFV11 To 2ISC*FT47X,FT488 215C*EFV13 To 215C*FT47H I 215C*EFV14- Vessel Bottom tap loop A Jet Pump l Inst. Line from M12,340' 215C*EFV15 l 2ISC*EFV17 Inst. Line from N12.20' .
- 215C*EFV18 To 2ISC*FT47J.FT48A 2ISC*EFV20 To 2ISC*FT47E 215C"EFV21 Vessel Bottom tap for C5il, RDS ""T"I 215C"EFV22 Vessel Botton Tap for WC5 and Loop 8 J.P.
2ISC*EFV23 To 2ISC*FT48C and Postaccident Sampling =
2ISC*EFV24 To 2ISC"FT480 and Postaccident Sampling l &
a 215C*EFV25 To 2ISC*FT47L l* 21st*EFV26 To 215C*FT47C 2ISC"EFV27 To 2ISC*FT47A C 2ISC*EFV28 To 2ISC*FT47R
-g 2ISC*EFV29 2ISC"EFV30 Io 2ISC*FI47G To 215C*F147N l ' " 215C*EFV31 To 215C*FT4BA i
- 2iSC*EFV32 To 2ISC*FT47T lg 215C"EFv33 To 215C"FT47V FI48C ,
\ ,
. ~ ...
! l ' }$$
1 %c
. TABLE 3.6.3-1 (Continued) 3
$ PRIMARY CONTAl MENT ISOLATION VALVES 5
N ~
, ISOLATION VALVE ISOLATION MAXIMUM CLO51HG o VALVE NO. VALVE FUNCTION GROUP SIGNAL (a) TIME- (SECONDS)
Y 2ISC"EFV34 To 2ISC*FT47B c 215C*EFV35 To 215t*FT47D E
~
215C*EFV36 l 10 2ISC"FT47F 2tSC"EFV37 To 215C"FT475 .
~ 2ISC*EFV38 To.215C*FT47M 2ISC"EFV39 To 2ISC*FT47P 215C*EFV40 To 215C"FT488 215C*EFV41 To 215C"FI47U 215C*EFV42 To 2ISC"FT4N,FT48D 2tSC*EFV9 Containment Pressure 2ISC*PT15C, 168, 16D u
s 2ISC*EFV12 Containment Pressure 2ISC'PT158,178,17D
- 215C*EFV16 Containment Pressure 2ISC*PT15A,16A,16C T 215C*EFV19 Containment Pressure 2ISC"Pil50,17A,17C w
2CM5"EFVIA To CMS *PTIA 2 CMS *EFV1B To CMS *PTIS ~
2 CMS *EFV3A To CMS *P12A 2 CMS'EFV38 To CMS *PT2B 2 CMS'EFVSA To CMS *PT7A 2 CMS *EFVS8 To CMS *PT78 2 CMS *EFV6 To CMS-Pil68 2 CMS *EFV8A To CMS *Li9A, llA, 114 *"E"Ig 2CM5"EFV88 To CMS *LT98, 118, 105 ,
2 CMS *EFV9A To CM5"LT9A, IIA, 114 2CM5"EFV98 2 CMS *EFV10 To CMS *LI98, 118, 105 Io CMS-PI173
)>
li"'""*"
2IC5"EFV1 To 2IC5"PD7167 C 21C5'EFV2 To 21CS*PDil67 - =g g 20ER"EFV31 To DER *PIl34 m
4
C'
. g g~ ;
-7, .
4 , ;, y g .g ,. 94
- - , . m. - .,
~g. .. . a ;;
c.
TABLE 3.6.3-1 (Continued)
PRIMARY CONTAIMENT ISOLATION VALVES .
3 r-
[ ISOLATION VALVE ISOLATION MAXIMUM CLOSING g VALVE NO. VALVE FUNCTION GROUP SIGNAL (a) TIME- (SECONOS) 5
, 2ICS*EFV3 To 21CS*PDT168 '
c 2ICS*EFV4 g To 2ICS*PDT168 x
% 2IAS*EFY200 To 2IAS*PT230 off ADS Acc m.
m 2IAS*EFV201 To 2IAS*PT231 off ADS Accum. ~
2IAS*EFV202 To 2IAS*PT232 of f ADS Accum.
2IAS*EFV203 To 2IAS*PT233 off ADS Accum.
2IAS*EFV204 To 2IAS*PT234 off ADS Accum. .
2fAS*EFV205 To 2IAS*PT235 off ADS Accum. .
2IAS*EFV206 To 2IAS*PT236 off ADS Accum.
5 2RHS*EFV 5. 6 To 2RHS*PDT188 -
2R11S*EFV7 To 2RHS*PDT18A 7
~
2 MSS *EFV 1A,B,C,0 To Flow elements A,B,C D steamlines 2MS$*EFV 2A,B.C.D To Flow elements A B.C.D steamlines 2 MSS *EFV 3A,8,C,0 To Flow elements A.B.C,0 steamlines '
2i IS*EFV 4A,B C.D To flow elements A B.C,0 steamlines 2RCS*EFV44 A,8 To 2RCS*PT 84 A/B 2RCS*EFV45 A,8 To .tCS*FT 7 A/8, FT 9 A/B -
2RCS*EFV46 A,8 2RCS*EFV4/ A,8 To J.;CS*FT 7 A/B, FT 9 A/8 To ;CS*FT 6 A/B, FT 8 A/B T'3 ,-
~
2RCS*EFV48 A,B To 1RCS*FT 6 A/8, FI 8 A/B 2RCS*EFV52 A,8 To 2RCS*PDT 15 A/8 D 2RCS*EFV53 A,B To 2RCS*PDT 15 A/B &
2RCS*EFV62 A,0 To 2RCS*PI44 A/B 2RCS*EFV63 A.E To 2RCS"PT42 A/B -
C h .
= .
21 a .
~
' ~
'!r ', '.'jfl l -
' l ':~.;- :'
. TABLE 3.6.3-1 (Continued) -
h PRIMARY CONTAINENT ISOLATION VALVES 1
! e-
'" VALVE ISOLATION MAXIMUM CLOSING IS0ggjon GROUP SIGNAt(a) TINE.(SECONDS)
$ VAD( W. VALVE FUNCTION
-x
" To 2WCS-FT 134
2WCS'gfV222 To 2WCS*FI67X, PDS 115 -
E 2WC98gfW223 I To 2WCS*FT67Y G 2WC$affv224 To 2WCS*FT67Y m 2WC$agfv300 To 2WCS*FT67X, PDS 115 2Cgyagfv1 To 2C5H*LT123, LT124 2Cgyagfy2 To 2CSM*LT123. LT124 .
2Cgyagfy3 A *PDI109 To 2C*
m ggfyl To 2CSL*PDi132 and 2RHS*PDil8A 2
t: -
< ~
D:=
-g b
Q E . ,
- NNAL UKAtl D '
TABLE 3.6.3-1 (Continued) ,.
PRIMARY CONTAINMENT ISOLATION VALVES
, TABLE NOTATION
(a) See Specification 3.3.2, Table 3.3.2-4, for valve groups operated by isolation signal (s).
(b) Deleted, l (c) These valves are the RHR heat exchangers vent lines isolation valves. The vent line connects to the RHR safety relief valves (SRVs) Discharge Header before it penetrates the primary containment. The position indicators for thesevalvesareprovidedintheCnntrol(Roomforremotemanualisolation.
(d) Type C leakage tests not required.
(e) The associated instrument lines shall not be isolated during Type A test- '
ing. Type C testing is not required These valves shall be tested in accordance with Surveillance Requitseent 4.6.3.4.
(f) These valves are check valves, located on the vacuum breaker lines for RHR SRVs discharge headers. The SRV discharge header terminates under pool water and therefore has no containment isolation yalv;s other than those on lines feeding into it.
(g) 2SLS*HOVSA and B are globe stop check valves. These valves close upon reverse flow. The motor operator is provided to remote manually close the ,
valve from the control room. '
l (h) These valves are tdstable check valves. They close upon reverse flow.
The air operator on each valve is provided only for periodic testing of the valve. These valves can only be tested against a zero d/p.
(i) Valves are. maintained close'd and the lines are capped. Valves are Type C tested.
(j) Not primary containment penetration isolation valves. These valves close on an isolation signal to provide integrity of "A" and "B" LPCI loops. l
. 1 (k) Valves close on a SCRAM signal; not part of primary containment isolation system but are included here for Type C testing per Specification 3.6.1.2.
These valves are not required to be OPERABLE per this specification but are required to be OPERABLE per Specification 3.1.3.1.
(1) Not subject to Type A or Type C leak test because of constant monitoring under constant-1800 psig pressure and the possible detrimental effects of shutdown. ,
(m) N'ot subject to Type C test per 10 C'FR 50, Appendix J. A hydr'ostatic test
. is performed in accordance with Specification 4.6.1.2.d.3.
(n) These valves are Type C tested in the reverse direction.
H!NE HILE POINT - UNIT 2 3/4 6-34 JUN25 x
.9
s i'
iffe: - *
,y'l
?
2"- SALP INPUT FROM THE PLANT SYSTEM BRANCH FOR NINE MILE POINT UNIT 2 PROPOSED b,. TECHNICAL SPECIFICATION CHANGES K'
- A. Licensing Activities
- f. 1. Management Involvement in Assuring Quality 4 During the review process the licensee's activities exhibited little evidence of prier planning.
Rating: 3
- 2. Approach to Resolution of Technical Issues from a Safety Standpoint.
During.the review some issues were not resolved in a timely ma'nner I
i Rating: 3
- 3. Responsiv'e to NRC Initiatives Rating: N/A
- 4. Staffing (includingManagement)
Rating: N/A
- 5. Reporting and Analysis of Reportable Events, Rating: N/A
, 6. Training and Qualification Effectiveness. ,
Rating: N/A ,
l
- 7. Overall r'ating for Licensing Activity Functical Area-1 Rating: 3 O
s 1
i l
\ _
_ - . .