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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217L8831999-10-21021 October 1999 Safety Evaluation Supporting Proposed Alternatives to Code Requirements Described in RR-V17 & RR-V18 ML20217G2041999-10-13013 October 1999 Safety Evaluation Supporting Amend 179 to License DPR-28 ML20212C2551999-09-17017 September 1999 Safety Evaluation Supporting Amend 175 to License DPR-28 ML20206E8741999-04-29029 April 1999 SER Determined That Flaw Evaluation Meets Rules of ASME Code & Assumed Crack Growth Rate Adequate for Application ML20205K7581999-04-0707 April 1999 Safety Evaluation Supporting Alternative Proposal for Reexamination of Circumferential Welds with Detected Flaw Indications in Plant RPV ML20203H9881999-02-18018 February 1999 SER Accepting Alternative to 10CFR50.55a(g)(6)(ii)(A) Augmented Reactor Vessel Exam at Vermont Yankee Nuclear Power Station.Technical Ltr Rept Encl ML20199K7151999-01-21021 January 1999 Corrected Safety Evaluation Supporting Amend 163 Issued to FOL DPR-28.Pages 2 & 3 Required Correction & Clarification ML20199K6991999-01-20020 January 1999 Safety Evaluation Concluding That Request to Use YAEC-1339, Yankee Atomic Electric Co Application of FIBWR2 Core Hydraulics Code to BWR Reload Analysis, at Vermont Yankee Acceptable ML20199L5951999-01-14014 January 1999 Safety Evaluation Accepting Licensee Proposed Alternative to Code Requirement,Described in Rev 2 to Pump Relief Request RR-P10 Pursuant to 10CFR50.55a(a)(3)(i) ML20195C4161998-11-0909 November 1998 SER Accepting Request That NRC Approve ASME Code Case N-560, Alternative Exam Requirement for Class 1,Category B-J Piping Welds ML20155B6471998-10-26026 October 1998 Safety Evaluation Accepting Jet Pump Riser Insp Results & Flaw Evaluation,Conducted During 1998 Refueling Outage ML20154B6951998-10-0101 October 1998 SER Re Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Vermont Yankee Nuclear Power Station ML20239A1361998-09-0202 September 1998 SER Re License Request for NRC Review & Concurrence W/Changes to NRC-approved Fire Protection Program ML20216F1001998-04-15015 April 1998 Safety Evaluation Accepting 980331 Licensee Proposal to Perform Alternative Testing for Containment Pressurization Test for Vynp ML20217F3421998-03-25025 March 1998 SER Accepting Plans for 1998 & 1999 Refueling Outages Re Reactor Vessel Internals for Plant ML20212H1521998-03-0606 March 1998 Correction to Page 7 of SE Re Relief Request for Third 10-yr Interval Pump & Valve IST Program for Plant ML20217N4911998-02-27027 February 1998 SER Pertaining to Cracking of EDG Lube Oil Piping at Vermont Yankee ML20198P9941998-01-15015 January 1998 SE Authorizing Relief Requests for Third Interval Pump & Valve Inservice Testing Program ML20141A4151997-06-18018 June 1997 Revised SE Accepting Proposed Onsite Disposal of Slightly Contaminated Silt Removed from Vermont Yankee Cooling Towers ML20135E5401997-03-0303 March 1997 Safety Assessment Accepting Mod of RHR & CS Sys Containment Isolation Function Configuration ML20134N8271996-11-20020 November 1996 Safety Evaluation Accepting Licensee Scope & Insp Methods Proposed for Insp of Core Spray Internal Piping During Fall 1996 Refueling Outage at Plant ML20134F9631996-11-0505 November 1996 Safety Evaluation Re Power/Flow Exclusion Region Calculation Method Using LAPUR5 Computer Code & Implementation of Solomon Stability Monitor for Licensee Facility ML20128N3531996-10-11011 October 1996 Safety Evaluation Accepting Licensee Flaw Evaluation of Indication Found During Reactor Pressure Vessel Insp at Plant ML20129G3611996-10-0202 October 1996 Safety Evaluation Accepting Proposed Repair for Plant Core Shroud ML20057A6991993-09-0303 September 1993 Safety Evaluation of IST Program Relief Requests for Pumps & Valves for Third 10-yr Insp Interval ML20057A2791993-08-12012 August 1993 Safety Evaluation Accepting Licensee Reasons Given for Delay in Completing short-term Actions Requested in Ieb 93-003, Resolution of Issues Re to Rv Water Level Instrumentation in Bwrs ML20246D7731989-08-21021 August 1989 Safety Evaluation Supporting Util Response to Generic Ltr 83-28,Item 2.2.1, Required Actions Based on Generic Implications of Salem ATWS Events. Equipment Classification Program for safety-related Components Acceptable ML20244D0311989-06-0707 June 1989 Safety Evaluation Accepting Util Second 10-yr Interval Inservice Insp Program Plan ML20205T4181988-10-14014 October 1988 Errata to Safety Evaluation Concluding Util Submittal Re Spent Fuel Pool Expansion ML20204F7271988-10-14014 October 1988 Safety Evaluation Supporting Proposed Expansion of Spent Fuel Pool at Facility ML20236N6461987-08-0707 August 1987 Safety Evaluation Re Permanent Elimination of Liquid Penetrant Exam of Feedwater Nozzles at Facility.Due to Lack of Reasonable Assurance That Ultrasonic Exam Can Totally Replace Penetrant Exam,Request Unacceptable ML20214T9891987-05-28028 May 1987 Safety Evaluation Re Util 870112 Proposed Plans to Inspect Two Overlay Repaired Core Spray safe-ends in Lieu of Replacement During Upcoming 1987 Refueling Outage.Plans Acceptable,Providing That Insp Results Satisfactory ML20207S7801987-03-12012 March 1987 Safety Evaluation Granting Relief from Tech Spec 4.7.A.3 on one-time Basis to Perform RHR Pump Wear Ring Replacement ML20214T4921986-11-24024 November 1986 Safety Evaluation Accepting Licensee 830511 & 860117 Responses to Generic Ltr 83-08 Re Mod of Vaccum Breakers on Mark I Containments ML20215M5871986-10-24024 October 1986 Preliminary Evaluation of Containment Study Transmitted w/860902 Ltr.Licensee Estimates Appear Optimistic Considering Uncertainties Inherent in Failure Rate Data ML20206F3651986-06-16016 June 1986 Safety Evaluation Re Proposed Repair of Core Spray safe- Ends,During Current Refueling Outage.Plant Can Be Safely Returned to Power Operation After Satisfactory Completion of Core Spray safe-end Repairs ML20206F0681986-06-13013 June 1986 Safety Evaluation Supporting 850514,0710,860327,0411 & 0513 Requests for Approval to Use Pvrc Damping Values (ASME Code Case N-411) for Piping Sys Reanalysis ML20202J4211986-03-31031 March 1986 Safety Evaluation Accepting Util Design Mods & Tech Spec Changes Re Degraded Grid Voltage Protection for Class 1E Sys.Lll Technical Evaluation Rept Encl ML20155B8351986-03-31031 March 1986 Safety Evaluation Supporting Revised Procedure OP-3140, Providing Technically Acceptable Actions During Degraded Grid Voltage Conditions W/O LOCA to Assure Protection of Class 1E Electrical Sys & Equipment ML20140H9881986-03-25025 March 1986 Safety Evaluation Re Util 851008 Request to Install Carpet Over Vinyl Asbestos Tiled Control Room Floor Covering. Installation of Carpet Will Not Decrease Level of Fire Safety in Control Room & Deviation Acceptable ML20138E4201985-12-0202 December 1985 Safety Evaluation Supporting Util 831107 & 840320 Responses 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 & Reactor Trip Sys Reliability ML20136F1241985-11-18018 November 1985 Safety Evaluation Re IE Bulletin 80-11, Masonry Wall Design. Issues Re Arching Action Theory Resolved ML20137S7331985-09-27027 September 1985 Safety Evaluation Approving Use of Fuel Thermal Performance Code,Frosstey,For Analysis of LOCA Conditions at Low & Moderate Burnups ML20135C8921985-09-10010 September 1985 Safety Evaluation Supporting 840824 Commitment to Convert Air Containment Atmosphere Dilution Sys to Nitrogen Sys,In Response to Generic Ltr 84-09 ML20135C9121985-09-10010 September 1985 Safety Evaluation Supporting Conclusion That Diversification of Scram Discharge Vol Level Instrumentation Not Necessary & Tech Specs,As Modified in Amend 76,resolve Staff Concerns Re Need for Instrumentation Diversity ML20134K7351985-08-19019 August 1985 Safety Evaluation Accepting 831107 Response to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Re post-maint Testing ML20136G3611985-08-12012 August 1985 Safety Evaluation Accepting Seismic Design Criteria Utilized for Evaluation of Modified Recirculation Sys ML20132D8971985-07-22022 July 1985 Safety Evaluation Supporting Use of Pvrc Damping Values (ASME Code Case N-411) for Response Spectrum Seismic Piping Analyses ML20127D8991985-05-0606 May 1985 Safety Evaluation Re 840925 & 1002 Responses to Generic Ltr 83-28,Item 1.1 Concerning post-trip Review Program & Procedures.Program & Procedures Acceptable 1999-09-17
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217L8831999-10-21021 October 1999 Safety Evaluation Supporting Proposed Alternatives to Code Requirements Described in RR-V17 & RR-V18 ML20217G2041999-10-13013 October 1999 Safety Evaluation Supporting Amend 179 to License DPR-28 BVY-99-127, Monthly Operating Rept for Sept 1999 for Vermont Yankee Nuclear Power Station.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Vermont Yankee Nuclear Power Station.With ML20212C2551999-09-17017 September 1999 Safety Evaluation Supporting Amend 175 to License DPR-28 BVY-99-112, Monthly Operating Rept for Aug 1999 for Vermont Yankee.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Vermont Yankee.With BVY-99-109, Ro:On 990812,stack Ng Effluent Instrumentation for PAM Was Declared Oos.Caused by Instrument Drift Due to Electronic Components Based on Insps by Instrumentation & Controls Dept.Detector & Preamplifier Will Be Replaced on 9908311999-08-19019 August 1999 Ro:On 990812,stack Ng Effluent Instrumentation for PAM Was Declared Oos.Caused by Instrument Drift Due to Electronic Components Based on Insps by Instrumentation & Controls Dept.Detector & Preamplifier Will Be Replaced on 990831 BVY-99-102, Monthly Operating Rept for July 1999 for Vermont Yankee. with1999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Vermont Yankee. with ML20209J0081999-07-14014 July 1999 Special Rept:On 990615,diesel Driven Fire Pump Failed to Achieve Rated Flow of 2500 Gallons Per Minute.Pump Was Inoperable for Greater than 7 Days.Corrective Maint Was Performed to Reset Pump Lift Setting BVY-99-090, Monthly Operating Rept for June 1999 for Vermont Yankee Nuclear Power Station.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Vermont Yankee Nuclear Power Station.With ML20196G5071999-06-23023 June 1999 Vynp Assessment of On-Site Disposal of Contaminated Soil by Land Spreading BVY-99-077, Monthly Operating Rept for May 1999 for Vermont Yankee Nuclear Power Station.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Vermont Yankee Nuclear Power Station.With BVY-99-068, Monthly Operating Rept for Apr 1999 for Vynp.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Vynp.With ML20206E8741999-04-29029 April 1999 SER Determined That Flaw Evaluation Meets Rules of ASME Code & Assumed Crack Growth Rate Adequate for Application ML20206D9301999-04-27027 April 1999 1999 Emergency Preparedness Exercise 990427 Exercise Manual (Plume Portion) ML20205S4211999-04-16016 April 1999 Non-proprietary Version of Revised Page 4-3 of HI-981932 Technical Rept for Vermont Yankee Spent Fuel Pool Storage Expansion ML20205K7581999-04-0707 April 1999 Safety Evaluation Supporting Alternative Proposal for Reexamination of Circumferential Welds with Detected Flaw Indications in Plant RPV BVY-99-046, Monthly Operating Rept for Mar 1999 for Vermont Yankee Nuclear Power Station.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Vermont Yankee Nuclear Power Station.With ML20205F6631999-03-0404 March 1999 Jet Pump Riser Weld Leakage Evaluation BVY-99-035, Monthly Operating Rept for Feb 1999 for Vermont Yankee Nuclear Station.With1999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Vermont Yankee Nuclear Station.With ML20205P8241999-02-28028 February 1999 Rev 2 to Vermont Yankee Cycle 20 Colr ML20203H9881999-02-18018 February 1999 SER Accepting Alternative to 10CFR50.55a(g)(6)(ii)(A) Augmented Reactor Vessel Exam at Vermont Yankee Nuclear Power Station.Technical Ltr Rept Encl ML20203A6951999-02-0404 February 1999 Revised Rev 2,App B to Vermont Yankee Operational QA Manual (Voqam) ML20199K7151999-01-21021 January 1999 Corrected Safety Evaluation Supporting Amend 163 Issued to FOL DPR-28.Pages 2 & 3 Required Correction & Clarification ML20199K6991999-01-20020 January 1999 Safety Evaluation Concluding That Request to Use YAEC-1339, Yankee Atomic Electric Co Application of FIBWR2 Core Hydraulics Code to BWR Reload Analysis, at Vermont Yankee Acceptable ML20199L5951999-01-14014 January 1999 Safety Evaluation Accepting Licensee Proposed Alternative to Code Requirement,Described in Rev 2 to Pump Relief Request RR-P10 Pursuant to 10CFR50.55a(a)(3)(i) BVY-99-071, Corp 1998 Annual Rept. with1998-12-31031 December 1998 Corp 1998 Annual Rept. with BVY-99-001, Monthly Operating Rept for Dec 1998 for Vermont Yankee Nuclear Power Station1998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Vermont Yankee Nuclear Power Station ML20198H5481998-12-23023 December 1998 Rev 2 to Vermont Operational QA Manual,Voqam ML20196H8641998-12-0101 December 1998 Cycle 19 Operating Rept BVY-98-163, Monthly Operating Rept for Nov 1998 for Vermont Yankee Nuclear Power Station.With1998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Vermont Yankee Nuclear Power Station.With ML20195C4161998-11-0909 November 1998 SER Accepting Request That NRC Approve ASME Code Case N-560, Alternative Exam Requirement for Class 1,Category B-J Piping Welds BVY-98-154, Monthly Operating Rept for Oct 1998 for Vermont Yankee Nuclear Power Station.With1998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Vermont Yankee Nuclear Power Station.With ML20155B6471998-10-26026 October 1998 Safety Evaluation Accepting Jet Pump Riser Insp Results & Flaw Evaluation,Conducted During 1998 Refueling Outage ML20154N0891998-10-16016 October 1998 Rev 1 to Vermont Operational QA Program Manual (Voqam) ML20154B6951998-10-0101 October 1998 SER Re Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Vermont Yankee Nuclear Power Station BVY-98-149, Monthly Operating Rept for Sept 1998 for Vermont Yankee Nuclear Power Station.With1998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Vermont Yankee Nuclear Power Station.With ML20239A1361998-09-0202 September 1998 SER Re License Request for NRC Review & Concurrence W/Changes to NRC-approved Fire Protection Program BVY-98-135, Monthly Operating Rept for Aug 1998 for Vermont Yankee Nuclear Power Station.With1998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Vermont Yankee Nuclear Power Station.With ML20151U0361998-08-28028 August 1998 Non-proprietary Rev 1 to Holtec Rept HI-981932, Vermont Yankee Nuclear Power Station Spent Storage Expansion Project ML20237E9221998-08-20020 August 1998 Vynp 1998 Form NIS-1 Owners Summary Rept for ISI, 961103-980603 BVY-98-122, Monthly Operating Rept for July 1998 for Vermont Yankee Nuclear Power Station1998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Vermont Yankee Nuclear Power Station ML20205F6491998-07-31031 July 1998 Rev 1 to GE-NE-B13-01935-02, Jet Pump Assembly Welds Flaw Evaluation Handbook for Vermont Yankee ML20236G0011998-06-30030 June 1998 Individual Plant Exam External Events BVY-98-098, Monthly Operating Rept for June 1998 for Vermont Yankee Nuclear Power Station1998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Vermont Yankee Nuclear Power Station ML20248C5081998-05-31031 May 1998 Rev 2 to 24A5416, Supplemental Reload Licensing Rept for Vermont Yankee Nuclear Power Station Reload 19 Cycle 20 ML20248C4951998-05-31031 May 1998 Rev 1 to Vermont Yankee Nuclear Power Station Cycle 20 Colr BVY-98-081, Monthly Operating Rept for May 1998 for Vermont Yankee Nuclear Power Station1998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Vermont Yankee Nuclear Power Station ML20247J8341998-05-31031 May 1998 Peak Suppression Pool Temp Analyses for Large Break LOCA Scenarios, for May 1998 ML20247G4001998-05-12012 May 1998 Interview Rept of Ej Massey ML20247E6351998-04-30030 April 1998 Rev 1 to GE-NE-B13-01935-LTR, Jet Pump Riser Welds Allowable Flaw Sizes Ltr Rept for Vermont Yankee 1999-09-30
[Table view] |
Text
._ _ .__ _ . _ _ _ _ . . _ . . . _ _ _ . _ _ . _ . _ _
i -
a mag g
p t UNITED STATES j
l g NUCLEAR REGULATORY COMMISSION
, 2 WASHINGTON, D.C. SegeMcM SAFETY ASSESSMENT BY THE'0FFICE OF NUCLEAR REACTOR REGULATION MODIFICATION OF THE CONTAIMENT ISOLATION CONFIGURATION OF THE CORE SPRAY AND RESIDUAL HEAT REMOVAL SYSTEMS VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET N0. 50-271
1.0 INTRODUCTION
Vermont Yankee Nuclear Power Corporation, has made changes to the isolation j valve configuration for the residual heat removal system and the core spray '
system of Vermont Yankee. These changes were made without prior NRC staff review and approval in accordance with 10 CFR 50.59. The following discussion documents the staff's subsequent assessment of these changes and of the licensee's review of these changes under 10 CFR 50.59 In 1973, the NRC amended 10 CFR Part 50 by adding Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Based on the requirements of Appendix J, it was concluded that Type C leak testing of these valves was not required because there were no postulated single active failures that would result in losing the 30-day water seals in the RHR and CS injection lines. This position was accepted by the NRC staff in its 1983 safety evaluation of Vermont Yankee's Appendix J 1eak rate testing program.
During a subsequent self assessment of the Vermont Yankee Appendix J 1eak rate testing program, VYNPC determined that Appendix J requirements were not being satisfied on several containment penetrations, including the CS injection line penetrations. This was reported to the NRC on March 1, 1996, in Licensee Event Report (LER) number 96-04 .
As a result, VYNPC installed test connections and attempted to perfom Type C tests on these check valves at the next opportunity during the fall 1996 refueling outage. However, these valves failed the Type C testing because they were not designed to seal at the relatively low pressure differentials used in Type C tests. In lieu of replacing these valves with valves designed to pass Type C test requirements, VYNPC proposed to modify the containment isolation configuration of the RHR and CS systems, pursuant to the change provisions of 10 CFR 50.59, prior to restart from the fall 1996 refueling outage.
During a series of telephone conversations with the NRC staff concluding on October 3,1996, representatives of VYNPC described changes planned in the containment isolation configuration for the RHR and CS systems prior to restart from the fall 1996 refueling outage. The NRC staff requested VYNPC 9703070131 970303 PDR ADOCK OS000271 p PDR
+
!- i i
tc submit a summary of VYNPC's evaluation of these changes. VYNPC submitted
- the requested information in a letter dated October 15, 1996. This letter was j supplemented on November 15, 1996. The following assessment of these changes
- by the NRC staff is based on the Sformation submitted in these letters.
! 2.0 RACKGROUND
- General Design Criteria of 10 CFR Part 50, Appendix A, address requirements i
for the containment isolation function of piping systems that penetrate primary containment. General Design Criterion (GDC) 55, " Reactor Coolant Pressure Boundary Penetrating Containment," defines acceptable configurations j of containment isolation valves (CIVs) for lines penetrating containment and i connected to the reactor coolant pressure boundary. In addition to listing l Specific acceptable containment isolation valve configurations, GDC 55 also provides for satisfying the containment isolation function "on some other i defined basis." t
! Because the RHR and the CS systems penetrate primary containment and are !
! connected to the reactor coolant pressure boundary, VYNPC applied GDC 55 of
- the General Design Criteria of 10 CFR Part 50, Aspendix A. The General Design
- Criteria are requirements for plants which have >een more recently licensed
- than Veraont Yankee. An acceptable alternative to satisfying GDC 55's "other
- defined basis" provision is a configuration consisting of a closed system and i one containment isolation valve outside primary containment for each '
l containment penetration. This alternative configuration is described in l ANSI /ANS-56.2 N271-1976, " Containment Isolation Provisions for Fluid Systems,"
which is endorsed by NRC Regulatory Guide (RG) 1.141, " Containment Isolation Provisions for Fluid Systems," dated April 1978. This standard was modified
, in 1984 but the changes do not affect any conclusions in this assessment.
t l 3.0 EVALUATION i The previous containment isolation configuration on each RHR and CS system i injection line, as described in the Vermont Yankee FSAR, consists of one i manually operated motor operated valve (MOV) outside containment and one check i valve inside containment. In addition, each injection line contains a non-CIV MOV. These valves are the following:
i System / MOVs (Outside Containment) Check Valves (Inside l Division Outboard Inboard CIV QILtainment)
RHR A RHR-25A RHR-27A RHR-46A
) RHR B RHR-25B RHR-27B RHR-46B
- CS A CS-IIA CS-12A CS-13A l CS B CS-118 CS-12B CS-138 i
i ;
i i
i i .
i i
1
- On a safety injection signal, both MOVs in each injection line receive a
- signal to open. The motor operated CIV has remote manual closure capability.
i However, the non-CIV motor operated valve is prevented from closing by the RHR
- or CS logic when a safety injection signal is present. In order to retain the
- capability to cope with a single failure of the containment isolation function ,
i of these systems, VYNPC has chosen to rely on a single motor operated valve on I i
each injection line and the capability of the RHR and CS systems outside j containment to function as closed systems.
3.1 Evaluation of CS and RHR Systems as Closed Systems 1
l VYNPC's evaluation cites GDC 55, " Reactor Coolant Pressure Boundary )
! Penetrating Containment" which permits a different CIV configuration based on l
! "some other defined basis" and ANS-56.2 ANSI N271-1976, " Containment Isolation Provisions for Fluid Systems," which is endorsed by NRC Regulatory Guide (RG)
- 1.141, " Containment Isolation Provisions for Fluid Systems", dated April 1978, )
which permits the use of one CIV outside containment and a closed system j
- outside containment. RG 1.141 provides the following criteria which must be
! satisfied in order to consider a system to be a closed system outside
! containment, i
j a. The closed system does not communicate with the outside atmosphere.
- b. The closed system meets Safety Class 2 design requirements.
! c. The closed system can withstand temperature and pressure equal to the j containment design conditions,
- d. The closed system can withstand the loss-of-coolant accident transient.
j e. The closed system meets seismic Category I design requirements.
I
- f. The closed system is protected against overpressure from thermal expansion of contained fluid when isolated, if required.
- g. The closed system is protected against high energy line breaks.
- h. The closed system is protected against loss of function from missiles.
- 1. The closed system is capable of being leak tested.
VYNPC's evaluation of the Vermont Yankee RHR and CS systems against each of the RG 1.141 closed-system criteria is contained in VYNPC's October ~15, 1996,
~
letter to the NRC. In its letter, VYNPC noted that the RHR and the CS systems )
are included in Vermont Yankee's leakage reduction program required by Vermont Yankee Technical Specification 6.10, " Integrity of Systems Outside Containment." Based on this and its review of VYNPC's evaluation, the NRC staff concludes that the Vermont Yankee RHR and CS systems outside containment may be considered closed systems as defined in RG 1.141. Although the RHR and CS check valves inside primary containment will no longer function as CIVs, '
i e f
these valves will remain operable with respect to all other required safety-
! related functions, such as pressure isolation. For these reasons, the change in containment isolation configuration for the RHR and CS systems is acceptable.
! 3.2 Evaluation for NRC Prior Approval i
VYNPC forwarded its 50.59 safety evaluation of this change by letter dated
! November 15, 1996. VYNPC applied the criteria in 10 CFR 50.59 to the change
}
and detemined that removing the CIV designation from the RHR and CS injection line check valves inside containment:
i 4
- a. Does not involve a change to the Vermont Yankee Technical Specifications (TSs) because none of these check valves are listed in TS Table 4.7.2a, I
" Primary Containment Isolation Valves - Valves Subject to Type C Leakage Tests."
- b. Does not involve an unreviewed safety question because:
(1) These check valves remain fully capable of performing all of their system and safety-related functions, other than containment isolation. In particular, these valves will continue to function as pressure isolation valves and be inspected and tested as specified by the applicable requirements of Section XI of the ASME Boiler and Pressure Vessel Code. Based on this, the NRC staff concludes that the change does not increase the probability of occurrence of an accident or malfunction of equipment important to safety, or create the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report.
(2) The previous and modified configurations are both acceptable, testable containment isolation valve configurations. The motor operated CIVs outside containment will still be subject to Type C Appendix J 1eakage rate testing. The RHR and CS systems outside containment meet the critaria of Regulatory Guide 1.141 for a closed system and are subject to leakage monitoring as required by the Vermont Yankee TS 6.10, " Integrity of Systems Outside Containment."
Therefore, the modified configuration is equivalent to the previous one. Based on this, the NRC staff concludes that the change does not result in an increase in radiological consequences or a reduction in the margin of safety as defined in the basis for any technical specification. . . ..
Based on the above, the NRC staff concludes that the RHR and CS system containment isolation function configuration change meets the criteria of 10 CFR 50.59 for changes that the licensee can make without prior Commission approval.
e
' ~
4 4
4.0 CONCLUSION
The NRC staff concludes that YYNPC's modification of the RHR and CS system
, containment isolation function configuration is acceptable because it provides a level of protection equivalent to the previous configuration. This equivalence is recognized in an industry standard which has been endorsed by an NRC regulatory guide, as discussed above. In addition, based on its review of VYNPC's submittals describing the change and the associated 50.59 safety evaluation, the NRC staff concludes that the change does not require prior NRC approval .
Principal Contributors: R. Lobel C. Harbuck Dated: March 3, 1997
. . -