NRC Generic Letter 1990-06

NRC Generic Letter 1990-006: Resolution of Generic Issue, Power-Operated Relief Valve and Block Valve Reliability, and Generic Issue 94, Additional Low-Temp. Overpressure Protection for Light-Water Reactors, Pursuant to 10 CFR 50354 (F)

ML031210416
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Washington Public Power Supply System, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Clinch River
Issue date:	06/25/1990
From:	Partlow J G Office of Nuclear Reactor Regulation
To:
References
GI-070, GI-094 GL-90-006, NUDOCS 9006200120
Download: ML031210416 (32)

'- /so UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D. C. 20555June 25, 1990TO: ALL PRESSURIZED WATER REACTOR LICENSEES AND CONSTRUCTIONPERMIT HOLDERSSUBJECT: RESOLUTION OF GENERIC ISSUE 70, "POWER-OPERATED RELIEFVALVE AND BLOCK VALVE RELIABILITY," AND GENERIC ISSUE 94,"ADDITIONAL LOW-TEMPERATURE OVERPRESSURE PROTECTION FORLIGHT-WATER REACTORS," PURSUANT TO 10 CFR 50.54(f)(GENERIC LETTER 90-06)The purpose of this generic letter is to advise pressurized water reactor (PWR)licensees and construction permit (CP) holders of the staff positions delineatedin Enclosures A and B to this letter. Enclosure A presents the staff positionresulting from the resolution of Generic Issue 70 (GI-70) and is applicable toall Westinghouse and Babcock and Wilcox (B&W)-designed plants and CombustionEngineering (CE)-designed plants with power-operated relief valves (PORVs).Enclosure B presents the staff position resulting from the resolution of GenericIssue 94 (GI-94) and is applicable to all Westinghouse-designed and CE-designedplants whether or not they have PORVs and block valves. Enclosure B does notapply to B&W-designed plants.The technical findings and the regulatory analysis related to GI-70 are discussedin NUREG-1316, "Technical Findings and Regulatory Analysis Related to GenericIssue 70--Evaluation of Power-Operated Relief Valve and Block Valve Reliabilityin PWR Nuclear Power Plants" (Enclosure C). In Enclosure D, the staff prepareda regulatory analysis for GI-94 based on the work performed by Battelle PacificNorthwest Laboratory (PNL) and reported in NUREG-1326, "Regulatory Analysis forthe Resolution of Generic Issue 94, Additional Low-Temperature OverpressureProtection for Light-Water Reactors."On the basis of technical studies for GI-70, the staff requests that to enhancesafety, actions identified in Section 3 of Enclosure A be taken by all PWRlicensees and CP holders that use or could use PORVs to perform any of thesafety-related functions identified in Section 2 of Enclosure A. These actionsresult from the staff interpretation of safety-related equipment (see 10 CFR50.49 and 10 CFR Part 100, Appendix A).On the basis of technical studies for GI-94, the staff also requests that toenhance safety, actions identified in Section 3 of Enclosure B be taken by allCombustion Engineering and Westinghouse PWR licensees and CP holders. Theseactions result from the staff interpretation of General Design Criteria 15 and31 in 10 CFR Part 50, Appendix A. The information requested by this letter isdirected at addressing these concerns.Note that the staff's requests are based on the performance of PORVs andPORV block valve designs used to date on U.S. power reactors. Currently,certain valve manufacturers are developing modified designs with the goal ofimproving reliability. The use of more reliable valves should result in lessfrequent corrective maintenance and can result in longer inservice testingintervals as delineated in Section XI of the ASME Boiler and Pressure VesselCode. >>r-I /"_200 _1*rtp

\ -'Generic Letter 90-06 -2 -Accordingly, pursuant to Section 182 of the Atomic Energy Act and 10 CFR 50.54(f),you, as a PWR licensee or CP holder, are required to advise the NRC staff underoath or affirmation, within 180 days of the date of this letter, of your currentplans relating to PORVs and block valves and to low-temperature overpressureprotection, in particular whether you intend to follow the staff positionsincluded in Enclosures A and B as applicable, or propose alternative measures,and your proposed schedule for implementation.For PWR plants with an operating license, staff positions 1 and 2 in Section 3.1of Enclosure A should be implemented by the end of the first refueling outagethat starts 6 months or later from the date of this letter. Requests for thetechnical specification modifications in staff position 3 in Section 3.1 ofEnclosure A and in Section 3 of Enclosure B should be submitted by the end ofthe first refueling outage that starts 6 months or later from the date of thisletter.For PWR CP holders, staff positions 1 and 2 in Section 3.1 of Enclosure A shouldbe implemented before initial criticality or within 6 months of the date ofthis letter, whichever is later. The technical specification modifications instaff position 3min Section 3.1 of Enclosure A and in Section 3 of Enclosure Bshould be submitted by the end of the first refueling outage that starts 6months or later from the date of this letter.If the applicable schedule cannot be met, the licensee or the CP holder shalladvise the staff of a proposed revised schedule, justification for any delay,and any planned compensating measures to be taken during the interim. Alterna-tives to schedules and the guidance provided herein will be evaluated on theirmerits on an individual case basis. Based on its review and the acceptabilityof these responses, the staff will close out GI-70 and GI-94 for each plant.Your response shall include the following specific items.1. A statement by licensees and CP holders as to whether they will commit toincorporate improvements 1, 2, and 3 in Section 3.1 of Enclosure A. Withrespect to improvement 3 in Section 3.1 of Enclosure A, licensees and CPholders shall state whether they will commit to use those modified limitingconditions of operation of PORVs and block valves in the technical specifica-tions for Modes 1, 2, and 3 in Attachment A-1 of Enclosure A for Westinghouse-designed and CE-designed plants with two PORVs, or in Attachment A-2 ofEnclosure A for Westinghouse-designed plants with three PORVs, or inAttachment A-4 of Enclosure A for B&W-designed plants. In addition tothis 10 CFR 50.54(f) request, if the licensees and the CP holders committo implement these recommended technical specifications, it is requestedthat they submit modifications to their current technical specificationsin a license amendment in accordance with the schedule noted above.1 Plants that already have staff-issued technical specifications consistentwith these requirements need merely state this in their response. No furtheraction will be required for this aspect of the Commission's position.

Generic Letter 90-06 -3 -2. A statement by licensees and CP holders as to whether they will submit alicense amendment request to modify the technical specifications and committo use the modified technical specifications for the low-temperatureoverpressure protection system concerning the limiting conditions ofoperation in Modes 5 and 6 as identified in Attachment B-1 of Enclosure Bto this generic letter for Westinghouse-designed or CE-designed plants, asappropriate. In addition to this 10 CFR 50.54(f) request, if the licenseesand CP holders commit to implement these recommended technical specifications,it is requested that they submit modifications to their current technicalspecifications in a license amendment in accordance with the schedule notedabove.The actions to incorporate technical specification (TS) requirements for theresolution of GI-70 and GI-94 are considered to be consistent with theCommission's Policy Statement on Technical Specification Improvements. Thispolicy statement captures existing requirements under Criterion 3 (Mitigationof Design-Basis Accidents or Transients) or under the provisions to retainrequirements that operating experience and probabilistic risk assessment showto be important to the public health and safety. Although it is recognizedthat PORVs for older plants may not have been classified as safety-relatedcomponents that are used to mitigate a design-basis accident and, therefore,may not have been included in TS as part of the plant's licensing basis, thisis not an acceptable basis for not implementing the proposed actions toincorporate TS requirements for PORVs consistent with the guidance provided.Likewise, such requirements would be retained in TS when implementing improve-ments in TS consistent with the Commission policy statement on the basis ofCriterion 3 or risk-considerations noted above.

Backfit Discussion

For GI-70, the actions proposed by the NRC staff to improve the reliability ofPORVs and block valves, as identified in Section 3 of Enclosure A, representnew staff positions for some licensees and CP holders, and this request isconsidered a backfit in accordance with NRC procedures. This backfit is acost-justified safety enhancement. Therefore, an analysis of the type describedin 10 CFR 50.109(a)(3) and 50.109(c) was performed, and a determination wasmade that there will be a substantial increase in overall protection of thepublic health and safety and that the attendant costs are justified in view ofthis increased protection. The analysis and determination will be made availablein the Public Document Room with the minutes of the 167th and 168th meetings ofthe Committee to Review Generic Requirements.It is noted that most of the recommended actions for GI-70 may already beimplemented by those plants that have received operating licenses in recentyears and would, therefore, represent less of a backfit than for older PWRplants that currently do not include PORVs and block valves in the ASMESection XI Inservice Testing Program and do not have technical specificationsfor PORVs and block valves or that operate with the block valves closedbecause of leaking PORVs.

I tGeneric Letter 90-06 -4 -For GI-94, the actions proposed by the NRC staff to improve the availabilityof the low-temperature overpressure protection (LTOP) system, as identified inSection 3 of Enclosure B, represent a new interpretation of existing requirementsfor some licensees and CP holders, and this request is considered a backfit inaccordance with NRC procedures. This backfit is a cost-justified safety enhance-ment. Therefore, an analysis of the type described in 10 CFR 50.109(a)(3) and50.109(c) was performed, and a determination was made that there will be asubstantial increase in overall protection of the public health and safety andthat the attendant costs are Justified in view of this increased protection-The analysis and determination will be made available in the Public DocumentRoom with the minutes of the 167th and 168th meetings of the Committee toReview Generic Requirements.This request is covered by Office of Management and Budget Clearance Number3150-0011, which expires January 31, 1991. The estimated average burden hoursis 320 person-hours per licensee response, including assessment of the newrecommendations, searching data sources, gathering and analyzing the data, andpreparing the required reports. Comments on the accuracy of this estimate andsuggestions to reduce the burden may be directed to the Office of Managementand Budget, Room 3208, New Executive Office Building, Washington, D.C. 20503,and the U.S. Nuclear Regulatory Commission, Information and Records ManagementBranch, Office of Information Resources Management, Washington, D.C. 20555.Sincerely,ats G. PartlowAssociate Director for ProjectsOffice of Nuclear Reactor Regulation

Technical Contact:

George A. Schwenk(301) 492-0878Enclosures:A. Staff Positions Resulting from Resolution of Generic Issue 70B. Staff Positions Resulting from Resolution of Generic Issue 94C. NUREG-1316, NTechnical Findings and Regulatory Analysis Relatedto Generic Issue 70--Evaluation of Power-Operated Relief Valve and BlockValve Reliability in PWR Nuclear Power Plants"D. NUREG-1326, "Regulatory Analysis for the Resolution of GenericIssue 94, Additional Low-Temperature Overpressure Protection for Light-Water Reactors"

eEnclosure A to Generic Letter 90-06Staff Positions Resulting fromResolution of Generic Issue 70 -PORV and Block Valve Reliability1.

BACKGROUND

Generic Issue 70 (GI-70), "Power-Operated Relief Valve and Block ValveReliability," involves the evaluation of the reliability of power-operatedrelief valves (PORVs) and block valves and their safety significance in PWRplants. The technical findings and regulatory analysis related to GI-70 arediscussed in NUREG-1316, "Technical Findings and Regulatory Analysis Relatedto Generic Issue 70--Evaluation of Power-Operated Relief Valve and BlockValve Reliability in PWR Nuclear Power Plants" (Enclosure C). This reportidentifies those safety-related functions that may be performed by PORVs andalso identifies potential improvements to PORVs and block valves. In supportof the resolution of GI-70, the Oak Ridge National Laboratory (ORNL) performeda study of PORV and block valve operating experience. A report, prepared byORNL, was issued as NUREG/CR-4692, "Operating Experience Review of Failures ofPower Operated Relief Valves and Block Valves in Nuclear Power Plants," datedOctober 1987.Traditionally, the PORV and its block valve are provided for plant operationalflexibility and for limiting the number of challenges to the safety-relatedpressurizer safety valves. The operation of the PORVs has not been classifiedas a safety-related function, i.e., one on which the results and conclusionsof the safety analysis are based and that invokes the highest level of qualityand construction. For overpressure protection of the reactor coolant pressureboundary (RCPB) at normal operating temperature and pressure, the operation ofPORVs has not been explicitly considered as a safety-related function. Also,an inadvertent opening of a PORV or safety valve has been analyzed in the FinalSafety Analysis Reports as an anticipated operational occurrence with acceptableconsequences. For these reasons, most PWRs, particularly those licensed priorto 1979, do not classify PORVs as safety-related components.The Three Mile Island Unit 2 (TMI-2) accident focused attention on the reliabilityof PORVs and block valves since the malfunction of the PORV at TMI-2 contributedto the severity of the accident. On other occasions, PORVs have stuck open whencalled upon to function. Also, there are PORVs in many operating plants thathave leakage problems so that the plants must be operated with the upstreamblock valves in the closed position. The technical specifications governingPORVs on most operating PWRs, which deal with closing the block valve andremoving power, were developed to allow continued plant operation with degradedPORVs, but did not consider the need for the PORVs to perform the safetyfunctions discussed below.Following the TMI-2 accident, the staff began to examine transient and accidentevents in more detail, particularly with respect to required operator actionsand equipment availability and performance. As a result, the staff initiatedan evaluation of the role of PORVs to perform certain safety-related functions.

'*1A-22. SAFETY FUNCTIONS OF PORVs AND BLOCK VALVESThe staff, in its evaluation, determined that over a period of time the roleof PORVs has changed such that PORVs are now relied upon by many Westinghouse,B&W, and CE designed plants with PORVs to perform one, or more, of the followingsafety-related functions:1. Mitigation of a design-basis steam generator tube rupture accident,2. Low-temperature overpressure protection of the reactor vessel duringstartup and shutdown, or3. Plant cooldown in compliance with Branch Technical Position RSB 5-1to SRP 5.4.7, "Residual Heat Removal (RHR) System."Where PORVs are used or could be used to perform one, or more, of thesafety-related functions identified above or to perform any other safety-relatedfunction that may be identified in the future, it is appropriate to reconsiderthe safety classification of PORVs and the associated block valves. For certainPWR plants receiving an operating license in recent years, the staff has requiredthese valves to be classified as safety-related components if they perform one,or more, safety-related functions.For operating PWR plants, the staff has concluded that it is not cost effectiveto replace (backfit) existing non-safety-grade PORVs and block valves (andassociated control systems) with PORVs and block valves that are safety gradeeven when they have been determined to perform any of the safety-relatedfunctions discussed above. Subsequent to the TMI-2 accident, a number ofimprovements were required of PORVs and block valves, such as requirements tobe powered from Class IE buses and to have valve position indication in thecontrol room. For operating plants, the greatest immediate benefits can bederived from implementing items 1 through 3 identified below, which can increasethe reliability of these components and provide assurance they will function asrequired.3. IMPROVEMENTS TO ALL PORVs AND BLOCK VALVES3.1 Operating PWR Plants and Construction Permit HoldersBased on the analysis and findings for GI-70, the staff concludes that thefollowing actions should be taken to improve the reliability of PORVs andblock valves:1. Include PORVs and block valves within the scope of an operationalquality assurance program that is in compliance with 10 CFR Part 50,Appendix B. This program should include the following elements:a. The addition of PORVs and block valves to the plant operationalQuality Assurance List.b. Implementation of a maintenance/refurbishment program for PORVs andblock valves that is based on the manufacturer's recommendations A-3or guidelines and is implemented by trained plant maintenancepersonnel.c. When replacement parts and spares, as well as complete components,are required for existing non-safety-grade PORVs and blockvalves (and associated control systems), it is the intent ofthis generic letter that these items may be procured inaccordance with the original construction codes and standards.2. Include PORVs, valves in PORV control air systems, and block valveswithin the scope of a program covered by Subsection IWV, "InserviceTesting of Valves in Nuclear Power Plants," of Section XI of the ASMEBoiler and Pressure Vessel Code. Stroke testing of PORVs should onlybe performed during Mode 3 (HOT STANDBY) or Mode 4 (HOT SHUTDOWN) andin all cases prior to establishing conditions where the PORVs areused for low-temperature overpressure protection. Stroke testing ofthe PORVs should not be performed during power operation. Additionally,the PORV block valves should be included in the licensees' expandedMOV test program discussed in NRC Generic Letter 89-10, "Safety-RelatedMotor Operated Valve Testing and Surveillance," dated June 28, 1989.3. For operating PWR plants, modify the limiting conditions of operation ofPORVs and block valves in the technical specifications for Modes 1,2, and 3 to incorporate the position adopted by the staff in recentlicensing actions. Attachments A-1 through A-3 are provided forguidance. The staff recognizes that some recently licensed PWR plantsalready have technical specifications in accordance with the staffposition. Such plants are already in compliance with this positionand need merely state that in their response. These recenttechnical specifications require that plants that run with the blockvalves closed (e.g., due to leaking PORVs) maintain electricalpower to the block valves so they can be readily opened from thecontrol room upon demand. Additionally, plant operation in Modes 1,2, and 3 with PORVs and block valves inoperable for reasons otherthan seat leakage is not permitted for periods of more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

-A-4 Generic Issue 70Enclosure A to Generic Letter 90-06Attachment A-1Modified Standard Technical Specificationsfor Combustion Engineering and Westinghouse PlantsREACTOR COOLANT SYSTEM3/4.4.4 RELIEF VALVESLIMITING CONDITION FOR OPERATIONThe following is to be used when two PORVs are provided:3.4.4 Both power-operated relief valves (PORVs) and their associated blockvalves shall be OPERABLE.APPLICABILITY: MODES 1, 2, and 3.ACTION:a. With one or both PORVs inoperable because of excessive seatleakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLEstatus or close the associated block valve(s) with power maintainedto the block valve(s); otherwise, be in at least HOT STANDBY withinthe next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.b. With one PORV inoperable due to causes other than excessive seatleakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status orclose its associated block valve and remove power from the blockvalve; restore the PORV to OPERABLE status within the following72 hours or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOTSHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.c. With both PORVs inoperable due to causes other than excessive seatleakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORV to OPERABLEstatus or close its associated block valve and remove power from theblock valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOTSHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.d. With one or both block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore theblock valve(s) to OPERABLE status or place its associated PORV(s) inmanual control. Restore at least one block valve to OPERABLE statuswithin the next hour if both block valves are inoperable; restoreany remaining inoperable block valve to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />;otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and inHOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

A-5 Generic Issue 70e. The provisions of Specification 3.0.4 are not applicable.SURVEILLANCE REQUIREMENTS4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORVshall be demonstrated OPERABLE at least once per 18 months by:a. Operating the PORV through one complete cycle of full travel duringMODES 3 or 4, and I .A-6 'G"eneric Issue 70Enclosure A To Generic Letter 90-06Attachment A-2Modified Standard Technical Specificationsfor Westinghouse Plants with Three PORVsREACTOR COOLANT SYSTEM3/4.4.4 RELIEF VALVESLIMITING CONDITION FOR OPERATIONThe following is to be used when three PORVs are provided:3.4.4 All power-operated relief valves (PORYs) and their associated blockvalves shall be OPERABLE.APPLICABILITY: MODES 1, 2, and 3.ACTION:a. With one or more PORVs inoperable because of excessive seat leakage,within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or closethe associated block valve(s) with power maintained to the blockvalve(s); otherwise, be in at least HOT STANDBY within the next 6hours and HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.b. With one or two PORVs inoperable due to causes other than excessiveseat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLEstatus or close the associated block valve(s) and remove power fromthe block valve(s); restore the PORY(s) to OPERABLE status within thefollowing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> andin HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.c. With three PORVs inoperable due to causes other than excessive seatleakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORV to OPERABLEstatus or close the block valves and remove power from the blockvalve(s) and be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOTSHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.d. With one or more block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore theblock valve(s) to OPERABLE status or place its associated PORV inmanual control. Restore at least one block valve to OPERABLE statuswithin the next hour if three block valves are inoperable; restoreany remaining inoperable block valve(s) to operable status within 72hours; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and inHOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

A-7Generic Issue 70e. The provisions of Specification 3.0.4 are not applicable.SURVEILLANCE REQUIREMENTS4.4.4.1 In addition to the requirements of Specification 4.0.5,shall be demonstrated OPERABLE at least once per 18 months by:each PORVa. Operating the PORV through one complete cycle of full travel duringMODES 3 or 4, andb. Where applicable, operating solenoid air control valves and checkvalves on associated air accumulators in PORV control systemsthrough one complete cycle of full travel for plants withair-operated PORVs, andc. Performing a CHANNEL CALIBRATION of the actuation instrumentation.4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92days by operating the valve through one complete cycle of full travel unless theblock valve is closed in order to meet the requirements of ACTION b, or c inSpecification 3.4.4.4.4.4.3 The emergency power supply for the PORVs and blockdemonstrated OPERABLE at least once per 18 months by:valves shall bea. Manually transferring motive andthe emergency power bus, andcontrol power from the normal tob. Operating the valves through a complete cycle of full travel.WESTINGHOUSE PLANTS

A-8Aeneric Issue 70Enclosure A to Generic Letter 90-06Attachment A-3Applicable to Combustion Engineering and Westinghouse Plants3/4.4.4 RELIEF VALVESBases of the Limiting Condition for Operation (LCO) and SurveillanceRequirements:The OPERABILITY of the PORVs and block valves is determined on the basis oftheir being capable of performing the following functions:A. Manual control of PORVs to control reactor coolant system pressure. Thisis a function that is used for the steam generator tube rupture accidentand for plant shutdown. This function has been classified as safetyrelated for more recent plant designs.B. Maintaining the integrity of the reactor coolant pressure boundary. Thisis a function that is related to controlling identified leakage andensuring the ability to detect unidentified reactor coolant pressureboundary leakage.C. Manual control of the block valve to: (1) unblock an isolated PORV toallow it to be used for manual control of reactor coolant system pressure(Item A), and (2) isolate a PORV with excessive seat leakage (Item B).D. Automatic control of PORVs to control reactor coolant system pressure.This is a function that reduces challenges to the code safety valves foroverpressurization events.E. Manual control of a block valve to isolate a stuck-open PORV.Surveillance Requirements provide the assurance that the PORVs and blockvalves can perform their functions. Specification 4.4.4.1 addresses PORVs,4.4.4.2 the block valves, and 4.4.4.3 the emergency (backup) power sources.The latter are provided for either PORVs or block valves, generally as aconsequence of the TMI ACTION requirements to upgrade the operability of PORVsand block valves, where they are installed with non-safety-grade powersources, including instrument air, and are provided with a backup (emergency)power source. The block valves are exempt from the surveillance requirementsto cycle the valves when they have been closed to comply with the ACTIONrequirements. This precludes the need to cycle the valves with full systemdifferential pressure or when maintenance is being performed to restore aninoperable PORV to operable status.Surveillance requirement 4.4.4.1.b has been added to include testing of themechanical and electrical aspects of control systems for air-operated PORVs.

->Generic Issue 70A-9Testing of PORVs in HOT STANDBY or HOT SHUTDOWN is required in order tosimulate the temperature and pressure environmental effects on PORVs. In manyPORV designs, testing at COLD SHUTDOWN is not considered to be a representativetest for assessing PORV performance under normal plant operating conditions.The Modified Standard Technical Specification (STS) requirements include thefollowing changes from prior STS guidance:1. Clarify the statement of LCO by replacing "All" with "Both" where the designincludes two PORVs. ' -2. ACTION statement a. includes the requirement to maintain power to closedblock valve(s) because removal of power would render the block valve(s)inoperable and the requirements of ACTION statement c. would apply. Power ismaintained to the block valve(s) so that it is operable and may be subsequentlyopened to allow the PORV to be used to control reactor pressure. Closure ofthe block valve(s) establishes reactor coolant pressure boundary integrity fora PORV that has excessive seat leakage. (Reactor coolant pressure boundaryintegrity takes priority over the capability of the PORV to mitigate anoverpressure event.) However, the APPLICABILITY requirements of the LCO tooperate with the block valve(s) closed with power maintained to the blockvalve(s) are only intended to permit operation of the plant for a limitedperiod of time not to exceed the next refueling outage (MODE 6) so thatmaintenance can be performed on the PORVs to eliminate the seat leakagecondition. The PORVs should normally be available for automatic mitigation ofoverpressure events and should be returned to OPERABLE status prior to enteringSTARTUP (MODE 2).3. ACTION statements b. and c. include the removal of power from a closed blockvalve as additional assurance to preclude any inadvertent opening of the blockvalve at a time in which the PORV may not be closed due to maintenance to restoreit to OPERABLE status. (In contrast, ACTION statement a. is intended to permitcontinued plant operation for a limited period of time with the block valvesclosed, i.e., continued operation is not dependent on maintenance at power toeliminate excessive PORV leakage, and, therefore, ACTION statement a. does notrequire removal of power from the block valve.)4. ACTION statements a., b., and c. have been changed to terminate the forcedshutdown requirements with the plant being in HOT SHUTDOWN rather than COLDSHUTDOWN because the APPLICABILITY requirements of the LCO do not extend pastthe HOT STANDBY mode.5. ACTION statement d. has been modified to establish remedial measures thatare consistent with the function of the block valves. The prime importance forthe capability to close the block valve is to isolate a stuck-open PORV. Therefore,if the block valve(s) cannot be restored to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, theremedial action is to place the PORV in manual control to preclude its automaticopening for an overpressure event and to avoid the potential for a stuck-openPORV at a time that the block valve is inoperable. The time allowed to restorethe block valve(s) to operable status is based upon the remedial action timelimits for inoperable PORVs per ACTION statements b. and c. since the PORVs_

A-10Generic Issue 70are not capable of mitigating an overpressure event when placed in manualcontrol. These actions are also consistent with the use of the PORVs to controlreactor coolant system pressure if the block valves are inoperable at a timewhen they have been closed to isolate PORVs that have excessive seat leakage.The modified ACTION statement does not specify closure of the block valvesbecause such action would not likely be possible when the block valve isinoperable. Likewise, it does not specify either the closure of the PORV,because it would not likely be open, or the removal of power from the PORV.When the block valve is inoperable, placing the PORV in manual control issufficient to preclude the potential for having a stuck-open PORY that couldnot be isolated because of an inoperable block valve. For the same reasons,reference is not made to ACTION statements b. and c. for the required remedialactions.6. Surveillance requirement 4.4.4.2 has been modified to remove the exceptionfor testing the block valves when they are closed to isolate an inoperable PORV.If the block valve is closed to isolate a PORV with excessive seat leakage, theoperability of the block valve is of importance, because opening the block valveis necessary to permit the PORV to be used for manual control of reactor pressure.If the block valve is closed to isolate an otherwise inoperable PORV, the maximumallowable outage time is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, which is well within the allowable limits (25percent) to extend the block valve surveillance interval (92 days). Furthermore,these test requirements would be completed by the reopening of a recently closedblock valve upon restoration of the PORV to operable status, i.e., completionof the ACTION statement fulfills the required surveillance requirement.

A-11Generic Issue 70Enclosure A to Generic Letter 90- 06Attachment A-4Modified Technical Specificationsfor Babcock and Wilcox PlantREACTOR COOLANT SYSTEM3/4.4.4 RELIEF VALVELIMITING CONDITION FOR OPERATION3.4.4 The power-operated relief valve (PORV) and its associatedshall be OPERABLE.block valveAPPLICABILITY:MODES 1, 2, and 3.ACTION:a. With the PORV inoperable because of excessive seat leakage, within 1hour either restore the PORV to OPERABLE status or close the associatedblock valve with power maintained to the block valve; otherwise, bein at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWNwithin the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.b. With the PORV inoperable due to causes other thanleakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV toclose the associated block valve and remove powerand be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> andthe following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.excessive seatOPERABLE status orfrom the block valve,in HOT SHUTDOWN withinc. With the block valve inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the blockvalves to OPERABLE status or place the associated PORV in manualcontrol and restore the block valve to operable status within thenext hour; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> andin HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.d. The provisions of Specification 3.0.4 are not applicable.SURVEILLANCE REQUIREMENTS4.4.4.1 In addition to the requirements of Specification 4.0.5, the PORV shallbe demonstrated OPERABLE at least once per 18 months by:a. Operating the PORV through one complete cycle of full travel duringMODES 3 or 4, andb. Performing a CHANNEL CALIBRATION of the actuation instrumentation.

A-12 Generic Issue 704.4.4.2 The block valve shall be demonstrated OPERABLE at least once per92 days by operating the valve through one complete cycle of full travel unlessthe block valve is closed in order to meet the requirements of ACTION b inSpecification 3.4.4.4.4.4.3 The emergency power supply for the PORV and block valve shall bedemonstrated OPERABLE at least once per 18 months by:a. Manually transferring motive and control power from the normal tothe emergency power bus, andb. Operating the valve through a complete cycle of full travel.BABCOCK & WILCOX PLANTS

.. IA-13 Generic Issue 70Enclosure A to Generic Letter 90-06Attachment A-5Applicable to Babcock and Wilcox Plants3/4.4.4 RELIEF VALVEBases of the Limiting Condition for Operation (LCO) and SurveillanceRequirements:The OPERABILITY of the PORV and block valve is determined on the basis oftheir being capable of performing the following functions:A. Manual control of the PORV to control reactor coolant system pressure.This is a function that is used for the steam generator tube ruptureaccident and for plant shutdown. This function has been classified assafety related for more recent plant designs.B. Maintaining the integrity of the reactor coolant pressure boundary. Thisis a function that is related to controlling identified leakage andensuring the ability to detect unidentified reactor coolant pressureboundary leakage.C. Manual control of the block valve to: (1) unblock an isolated PORV toallow it to be used for manual control of reactor coolant system pressure(Item A), and (2) isolate the PORV with excessive seat leakage (Item B).D. Automatic control of the PORV to control reactor coolant system pressure.This is a function that reduces challenges to the code safety valves foroverpressurization events.E. Manual control of a block valve to isolate a stuck-open PORV.Surveillance Requirements provide the assurance that the PORV and blockvalve can perform their functions. Specification 4.4.4.1 addresses the PORV,4.4.4.2 the block valve, and 4.4.4.3 the emergency (backup) power source.The latter is provided for either the PORV or block valve, generally as aconsequence of the TMI ACTION requirements to upgrade the operability of PORYsand block valves, where they are installed with non-safety-grade powersources, including instrument air, and are provided with backup (emergency)power sources. The block valve is exempt from the surveillance requirementsto cycle the valve when it has been closed to comply with the ACTIONrequirements. This precludes the need to cycle the valve with full systemdifferential pressure or when maintenance is being performed to restore aninoperable PORY to operable status.

A-14Generic Issue 70Testing the PORV in HOT STANDBY or HOT SHUTDOWN is required in order tosimulate the temperature and pressure environmental effects on the PORV. Inmany PORV designs, testing at COLD SHUTDOWN is not considered to be arepresentative test for assessing PORV performance under normal plant operatingconditions.The Modified Standard Technical Specification (STS) requirements include thefollowing changes from prior STS guidance:1. ACTION statement a. includes the requirement to maintain power to the closedblock valve, because removal of power would render the block valve inoperableand the requirements of ACTION statement c. would apply. Power is maintainedto the block valve so that it is operable and may be subsequently opened toallow the PORV to be used to control reactor pressure. Closure of the blockvalve establishes reactor coolant pressure boundary integrity for a PORV thathas excessive seat leakage. (Reactor coolant pressure boundary integrity takespriority over the capability of the PORY to mitigate an overpressure event.)However, the APPLICABILITY requirement of the LCO to operate with the blockvalve closed with power maintained to the block valve is only intended to permitoperation of the plant for a limited period of time not to exceed the nextrefueling outage (MODE 6) so that maintenance can be performed on the PORV toeliminate the seat leakage condition. The PORV should normally be available forautomatic mitigation of overpressure events and should be returned to OPERABLEstatus prior to entering STARTUP (MODE 2).2. ACTION statement b. includes the removal of power from the closed blockvalve as additional assurance to preclude any inadvertent opening of the blockvalve at a time in which the PORV may not be closed due to maintenance to restoreit to OPERABLE status. (In contrast, ACTION statement a. is intended to permitcontinued plant operation for a limited period of time with the block valveclosed, i.e., continued operation is not dependent on maintenance at power toeliminate excessive PORV leakage, and, therefore, ACTION statement a. does notrequire removal of power from the block valve.)3. ACTION statements a. and b. have been changed to terminate the forced shutdownrequirements with the plant being in HOT SHUTDOWN rather than COLD SHUTDOWNbecause the APPLICABILITY requirements of the LCO do not extend past the HOTSTANDBY mode.4. ACTION statement c. has been modified to establish remedial measures thatare consistent with the function of the block valves. The prime importance forthe capability to close the block valve is to isolate a stuck-open PORV.Therefore, if the block valve cannot be restored to operable status within 1hour, the remedial action is to place the PORV in manual control to precludeits opening for an overpressure event and to avoid the potential for a stuck-open PORV at a time that the block valve is inoperable. The time allowed torestore the block valve to operable status is based upon the remedial actiontime limits for an inoperable PORV per ACTION statement b. since the PORV is notcapable of mitigating an overpressure event when placed in manual control.This action is also consistent with the use of the PORV to control reactorcoolant system pressure if the block valve is inoperable at a time when it was A-15Generic Issue 70closed to isolate a PORV that has excessive seat leakage. The modified ACTIONstatement does not specify closure of the block valve because such actionwould not likely be possible when the block valve is inoperable. Likewise, itdoes not specify either the closure of the PORV, because it would not likelybe open, or the removal of power from the PORV. When the block valve isinoperable, placing the PORV in manual control is sufficient to preclude thepotential for having a stuck-open PORV that could not be isolated because of aninoperable block valve. For the same reasons, reference is not made to ACTIONstatement b. for the required remedial action.5. Surveillance requirement 4.4.4.2 has been modified to remove the exceptionfor testing the block valve when it is closed to isolate an inoperable PORV.If the block valve is closed to isolate a PORV with excessive seat leakage, theoperability of the block valve is of importance, because opening the block valveis necessary to permit the PORV to be used for manual control of reactor pressure.If the block valve is closed to isolate an otherwise inoperable PORV, the maximumallowable outage time is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, which is well within the allowable limits (25percent) to extend the block valve surveillance interval (92 days). Furthermore,these test requirements would be completed by the reopening of a recently closedblock valve upon restoration of the PORV to operable status, i.e., completionof the ACTION statement fulfills the required surveillance requirement.

B-1Enclosure B to Generic Letter 90-06Staff Positions Resulting fromResolution of Generic Issue 94 -Additional Low-Temperature Overpressure ProtectionFor Light-Water Reactors1.

BACKGROUND

Generic Issue 94 (GI-94), "Additional Low-Temperature Overpressure Protectionfor Light-Water Reactors," addresses concerns with the implementation of therequirements set forth in the resolution of Unresolved Safety Issue (USI) A-26,"Reactor Vessel Pressure Transient Protection (Overpressure Protection)." Insupport of GI-94, the Battelle Pacific Northwest Laboratories (PNL) performed astudy based on actual operating reactor experiences to determine the risksassociated with current low-temperature overpressure protection (LTOP) systems.A report, prepared by PNL, has been issued as NUREG/CR-5186, "Value/ImpactAnalysis of Generic Issue 94, Additional Low Temperature Overpressure Protectionfor Light-Water Reactors," dated November 1988. The staff has prepared aregulatory analysis for GI-94 based on the work performed by PNL and reportedin NUREG-1326, "Regulatory Analysis for the Resolution of Generic Issue 94,Additional Low-Temperature Overpressure Protection for Light-Water Reactors"(Enclosure D).Low-temperature overpressure protection (LTOP) was designated as UnresolvedSafety Issue A-26 in 1978 (NUREG-0371). PWR licensees implemented proceduresto reduce the potential for overpressure events and installed equipmentmodifications to mitigate such events based on the staff recommendations fromthe USI A-26 evaluations, under Multi-Plant Action Item B-04 (NUREG-0748).Current staff guidelines for LTOP are in Standard Review Plan Section 5.2.2,"Overpressure Protection," and in its attached Branch Technical Position (BTP)RSB 5-2, "Overpressure Protection of Pressurized Water Reactors While Operatingat Low Temperatures" (NUREG-0800).The administrative controls and procedures that were identified as part ofMulti-Plant Action Item B-04 include the following items:1. Minimize the time the reactor coolant system (RCS) is maintained in awater-solid condition.2. Restrict the number of high-pressure safety injection pumps operableto no more than one when the RCS is in the LTOP condition.3. Ensure that the steam generator to RCS temperature difference is lessthan 50 Deg F when a reactor coolant pump (RCP) is being started in awater-solid RCS.4. Set the PORV setpoint (if the particular plant relies on thiscomponent for LTOP) to a plant-specific analysis supported value, andhave surveillance that checks the PORV actuation electronics and setpoint.

B-2Twelve PWR overpressure transients were reported during the period from 1981 to1983 after completion of USI A-26. Two of these events, at Turkey Point Unit4, exceeded the pressure/temperature limits of the technical specifications.During this same timeframe, there were 37 reported instances when at least oneLTOP channel was out of service. In 12 of these cases, both LTOP channels wereinoperable.The continuation of overpressure transient events, and the unavailability ofLTOP protection channels, suggested the need to reevaluate the currentoverpressure protection requirements, or their implementation, to determinewhether additional measures are warranted.Major overpressurization of the reactor coolant system while at lowtemperature, if combined with a critical crack in the reactor pressure vesselwelds or plate material, could result in a brittle fracture of the pressurevessel. Failure of the pressure vessel could make it impossible to provideadequate coolant to the reactor core and result in major core damage or a coremelt accident.The safety significance of these continuing low-temperature overpressuretransients was designated as Generic Issue 94, "Additional Low TemperatureOverpressure Protection." The concerns of GI-94 are applicable to all PWRplants regardless of the features used to mitigate a low-temperatureoverpressure event or of any measures to preclude events that would challengethese features or exceed the design basis for LTOP.The implementation of the requirement for an LTOP system (the resolution ofUSI A-26) has been found to be essentially uniform for the Combustion Engineering(CE) and Westinghouse (W) PWRs. With the exception of a few plants,* the LTOPprotection systems consist of either redundant PORVs or redundant safety reliefvalves (SRVs) in the residual heat removal (RHR) system and in general meet theguidance set forth in Branch Technical Position RSB 5-2, "OverpressurizationProtection of Pressurized Water Reactors While Operating at Low Temperatures."Variability in meeting IEEE-279 requirements, equipment environmentalqualification, and in meeting the guidance of Regulatory Guide 1.26, "QualityGroup Classification and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants," exists. As part of the NRC staffacceptance of LTOP protection system designs for the implementation of theresolution of USI A-26, it was concluded that the costs associated with upgradingexisting systems to meet the guidance of Regulatory Guide 1.26 were not* CE -San Onofre Units 2 and 3 rely on a single RHR (SDCS) SRV for LTOP.With the SRV inoperable, depressurize and vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.-Maine Yankee relies on two PORYs when pressure is above 400 psigand two RHR SRVs when pressure is below 400 psig.W -DC Cook Units 1 and 2 rely on either two PORVs or one PORV and oneRHR SRV.-Yankee Rowe relies on one PORV and two RHR SRVs.-Newer Westinghouse plants allow either two PORVs or two RHR SRVs.

B-3justifiable. Further evaluations performed for GI-94 have also concluded thatit is not cost beneficial to upgrade these systems to fully safety-gradestandards.2. CURRENT STANDARD TECHNICAL SPECIFICATION REQUIREMENTSThe section of the Standard Technical Specifications (STS) covering the LTOPprotection system is entitled Overpressure Protection System, Section 3.4.10.3for CE plants and Section 3.4.9.3 for W plants. The LTOP system setpoints areestablished based on additional restrictions for the restart of an idle reactorcoolant pump and on the number of high-pressure safety injection pumps and/orcoolant charging pumps allowed to be operable when LTOP is required. Theseadditional restrictions define the initial conditions for the plant-specifictransient analyses performed to establish the LTOP system setpoints. Theadditional restrictions are provided regarding the restart of inactive reactorcoolant pumps in Sections 3.4.1.3 (Hot Shutdown) and 3.4.1.4 (Cold Shutdown).High-pressure safety injection pump operability restrictions are provided inSection 3/4.5.3 (ECCS Subsystems). In addition to these administrativerestrictions, the transient analyses are based on a dual-channel system beingoperable to satisfy the single failure criterion of 10 CFR Part 50, Appendix A,for a system that performs a safety function. Therefore, the OverpressureProtection System TS is consistent with Criterion 2 of the Commission's PolicyStatement on Technical Specification Improvements for Nuclear Power Plants.The TS also satisfied Criterion 3 of the policy statement in that the LTOPsystem is the primary success path for the mitigation of low-temperatureoverpressure transients that present a challenge to a fission product barrier,in this case, the reactor pressure vessel.PORVs are relied on, by most Westinghouse designed plants and about one-half ofthe Combustion Engineering plants, to provide LTOP protection. In addition toPORMs, the RHR SRVs are also relied on to provide LTOP protection for some Wplants and for the CE plants that do not have PORVs. Newer W plants have TSthat require either two PORVs or two RHR SRVs for LTOP protection.The current STS ACTION requirements for the LTOP system include a 7-dayallowable outage time (AOT) to restore an inoperable LTOP channel to operablestatus before other remedial measures would have to be taken. In addition,ACTION d. states that the provisions of Specification 3.0.4 are not applicable.Therefore, the plant may enter the modes for which the Limiting Conditions forOperation (LCO) apply, during a plant shutdown or placement of the head onthe vessel following refueling, when an LTOP channel is inoperable. In thissituation, the 7-day AOT applies for restoring the channel to operable statusbefore other remedial measures would have to be taken. This is the same mannerin which the ACTION requirements apply when an LTOP channel is determined to beinoperable while the plant is in a mode for which the LTOP system is requiredto be operable.Based on the NRC evaluation of the LTOP system unavailability, it is concludedthat additional restrictions on' operation with an inoperable LTOP channel arewarranted when the potential for a low-temperature overpressure event is the B-4highest, and especially when the plant is in a water-solid condition.Furthermore, it is concluded that the additional restrictions regarding therestart of inactive reactor coolant pumps and regarding the operability of high-pressure safety injection pumps should be implemented in the TS, as indicatedin the STS, and licensees should verify that these administrative restrictionshave been implemented. Finally, it is concluded that these additional measureswill help to emphasize the importance of the LTOP system, especially whileoperating in a water-solid condition, as the primary success path for themitigation of overpressure transients during low-temperature operation.3. IMPROVEMENTS IN PROTECTION SYSTEM AVAILABILITYThe staff has determined that LTOP protection system unavailability is thedominant contributor to risk from low-temperature overpressure transients. Thestaff has further concluded that a substantial improvement in availabilitywhen the potential for an overpressure event is the highest, and especiallyduring water-solid operations, can be achieved through improved administrativerestrictions on the LTOP system.In developing the staff position on the resolution of the low-temperatureoverpressure protection generic issue, a number of factors have been taken intoconsideration.The staff has considered the conditions under which a low-temperatureoverpressure transient is most likely to occur. While LTOP protection isrequired for all shutdown modes, the most vulnerable period of time was foundto be MODE 5 (Cold Shutdown) with the reactor coolant temperature less than orequal to 200 Deg F, especially when water-solid, based on the detailed evaluationof operating reactor experiences performed in support of GI-94. LTOP transientsthat have challenged the overpressure protection system have occurred withreactor coolant temperatures in the range of 80 Deg F to 190 Deg F. In addition,a review of the STS for containment integrity indicates that there are nospecific requirements imposed during MODE 5, when the reactor coolant temperatureis below 200 Deg F. Industry responses to Generic Letter 87-12, "Loss of RHRWhile RCS Partially Filled," dated July 9, 1987, also indicate that containmentintegrity during MODE 5 is often relaxed to allow for testing, maintenance, andthe repair of equipment.In addition, the staff takes note of the fact that, in all instances whenpressure/temperatures limits in the TS have been exceeded, one LTOP protectionchannel was removed from service for maintenance-related activities. Duringthese events the redundant LTOP protection channel failed to mitigate theoverpressure transient as a result of a system/component failure that had notbeen detected.The reported LTOP transients have occurred in MODE 5 with RCS temperaturesranging from 80 Deg F to 190 Deg F. Since this temperature range includesMODE 6, RCS temperature less than 140 Deg F but with k-eff less than 0.95 ascompared to k-eff less than 0.99 for MODE 5, the staff concludes that theadditional administrative restriction for the single channel AOT is applicableto MODE 5 and MODE 6 (with the reactor pressure vessel head on).

B-5The staff concludes that the LTOP system performs a safety-related function andinoperable LTOP equipment should be restored to an operable status in a shorterperiod of time. The current 7-day AOT for a single channel is consideredto be too long under certain conditions. The staff has concluded that the AOTfor a single channel should be reduced to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when operating in MODE 5or 6 when the potential for an overpressure transient is highest. Theoperating reactor experiences indicate that these events occur during plannedheatup (restart of an idle reactor coolant pump) or as a result of maintenanceand testing errors while in MODE 5. The reduced AOT for a single channel inMODES 5 and 6 will help to emphasize the importance of the LTOP system inmitigating overpressure transients and provide additional assurance that plantoperation is consistent with the design basis transient analyses.Based on the foregoing concerns, added assurance of LTOP availability is to beprovided by revising the current Technical Specification for OverpressureProtection to reduce the AOT for a single channel from 7 days to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> whenthe plant is operating in MODES 5 or 6. Attachment B-1 is provided for guidancefor Westinghouse and CE plants. The guidance provided is also applicable toplants that rely on both PORVs and RHR SRVs or that rely on RHR SRVs only.Attachment B-2 provides the staff bases for the Overpressure ProtectionTechnical Specification.In performing the studies for GI-94, the staff has assumed that theadministrative controls and procedures identified in Section 1 have beenimplemented to ensure that the plant is being operated within the design base.If it is determined that the design base was developed based on restrictedsafety injection pump operability and/or differential temperature restrictionsfor RCP restart and that these restrictions have not been implemented as partof USI A-26, then these restrictions should be implemented now. This is not anew requirement. Attachment B-3 is provided for guidance.

B-6Generic Issue 94Enclosure B to Generic Letter 90-06Attachment B-1Modified Technical Specificationsfor Combustion Engineering and Westinghouse PlantsREACTOR COOLANT SYSTEMOVERPRESSURE PROTECTION SYSTEMLIMITING CONDITION FOR OPERATION3.4.9.3 Two power-operated relief valves (PORVs) shall be OPERABLE with alift setting of less than or equal to [450] psig.APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less thanor equal to 1275]OF, MODE 5, and MODE 6 when the head is on the reactor vesseland the RCS is not vented through a _ square inch or larger vent.ACTION:a. With one PORV inoperable in MODE 4, restore the inoperable PORV toOPERABLE status within 7 days or depressurize and vent the RCSthrough at least a square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.b. With one PORV inoperable in MODES 5 or 6, either (1) restore theinoperable PORV to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or (2) completedepressurization and venting of the RCS through at least a squareinch vent within a total of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />.c. With both PORVs inoperable, complete depressurization and venting ofthe RCS through at least a square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.d. With the RCS vented per ACTIONS a, b, or c, verify the vent pathwayat least once per 31 days when the pathway is provided by a valve(s)that is locked, sealed, or otherwise secured in the open position;otherwise, verify the vent pathway every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.e. In the event either the PORVs or the RCS vent(s) are used to mitigatean RCS pressure transient, a Special Report shall be prepared andsubmitted to the Commission pursuant to Specification 6.9.2 within30 days. The report shall describe the circumstances initiating thetransient, the effect of the PORVs or RCS vent(s) on the transient,and any corrective action necessary to prevent recurrence.f. The provisions of Specification 3.0.4 are not applicable.

B-7 Generic Issue 94SURVEILLANCE REQUIREMENTS4.4.9.3 Each PORV shall be demonstrated OPERABLE by:a. Performance of an ANALOG CHANNEL OPERATIONAL TEST, but excludingvalve operation, at least once per 31 days; andb. Performance of a CHANNEL CALIBRATION at least once per 18 months; andc. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

B-8Generic Issue 94Enclosure B to Generic Letter 90-Attachment B-23/4.4,9.3 OVERPRESSURE PROTECTION SYSTEMBases of the Limiting Condition for Operation and Surveillance Requirements:The OPERABILITY of the PORVs is determined on the basis of their being capableof performing the function to mitigate an overpressure event during low-temperature operation.The Modified Standard Technical Specification (STS) requirements include thefollowing changes from prior STS guidance:1. The depressurizing and venting of the RCS is not classified as anoverpressure protection system. However, the APPLICABILITY of the LCOexcludes MODE 6 when the RCS is adequately vented. This avoids anypossible question on Specification 3.0.4 being applied to preclude placementof the head on the vessel if any part of the LCO is not met when the RCSis vented.2. The APPLICABILITY for MODE 6 is clarified as "when the head is on thereactor vessel" rather than as "MODE 6 with the reactor vessel head on."3. ACTION a. is revised to clarify that it is only applicable in MODE 4.4. ACTION b. was added to reduce the allowed outage time for aninoperable PORV to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in MODES 5 or 6. Because this LCO does notapply under certain conditions specified under the APPLICABILITY for thisspecification, the ACTION statements likewise do not apply under thoseconditions. ACTIONS a. and b. do not repeat those qualifying conditionsthat apply for these modes since the actions only apply when the unit isunder those conditions.5. ACTION d. includes the requirements to verify that ACTIONS a., b., orc. continue to be met on an ongoing basis when the unit would be in MODES 4,5, or 6.6. The Surveillance Requirements were simplified by removing requirementsthat exist because of the general requirements applicable to all surveillancerequirements as specified in Section 4.0 of the TS.7. Surveillance Requirement 4.4.9.3.2 was removed since it is addressedby ACTION d.For plants with existing TS for PORVs used for LTOP, the only required changeis that indicated to restrict the applicability of ACTION a. to MODE 4 and forincorporating ACTION b. Any other changes that are proposed consistent with B-9 Generic Issue 94the above guidance are voluntary. For a plant without existing TS for PORVsthat are used for LTOP, a TS should be proposed that conforms to the aboveguidance.Becaus~e some plants use residual heat removal (RHR) safety relief valves forLTOP, either in addition to or in lieu of PORVs, similar requirements areincluded in TS as noted above for PORVs, The same changes in ACTION requirementsa. and b. are required, as noted above, for these plants. Likewise, any plantwithout existing TS for RHR suction relief valves that are used for LTOP shouldpropose TS that are consistent with the above guidance. When only RHR safetyrelief valves are used for LTOP, the Surveillance Requirements would state: "Noadditional requirements other than those required by Specification 4.0.5."

-B-10Generic Issue 94 -Enclosure B to Generic Letter 90-Attachment B-3Technical Specifications Guidancefor Combustion Engineering and Westinghouse PlantsOperational Limitations Consistent With the Design Basis Assumptions for theLow-temperature Overpressure Protection (LTOP) SystemThe TS requirements for LTOP typically apply in MODE 4 when the temperature ofany cold leg is below 2750F, MODE 5, and MODE 6 when the head is on thereactor vessel. During these conditions, one train (or channel) of the LTOPsystem is capable of mitigating an LTOP event that is bounded by the largestmass addition to the RCS or by the largest increase in RCS temperature that canoccur. The largest mass addition to the RCS is limited based upon the assumptionthat no more than a fixed number of pumps are capable of providing makeup orinjection into the RCS. Hence, this is a matter important to safety that pumpsin excess of this design basis assumption for LTOP not be capable of providingmakeup or injection to the RCS.The capability for makeup and injection to the RCS is also a safety concernfor normal makeup to the reactor coolant system for reactivity control as wellas for events that could result in a loss of coolant from the RCS. Theformer are covered by Technical Specifications (TS) under Reactivity ControlSystems, Charging Pump -Shutdown (MODES 5 and 6); Charging Pumps -Operating(MODES 1 through 4); and Flow Paths -Operating (MODES 1 through 4). Thelatter is covered by TS under Emergency Core Cooling Systems, ECCS Subsystems -Tcold Less Than 3500F (MODE 4).The manner in which restrictions, consistent with the design basis assumptionsof the LTOP system, have been incorporated in TS that require the operabilityof makeup or injection pumps has varied depending upon plant-specificconsiderations for the LTOP design and plant-specific designs for the use ofpumps for makeup and ECCS functions. A common method has been the use offootnotes to the pump operability requirements to note that:A maximum of one Safety Injection [and/or] one charging pump shall beOPERABLE when the temperature of one or more of the RCS cold legs is lessthan 2750F.This footnote is used for each specification that requires the operability ofa safety injection and/or charging pump in MODES 4 or 5.The Surveillance Requirements typically include the following:All Safety Injection [and/or] charging pumps, except the above requiredOPERABLE pump[s], shall be demonstrated to be inoperable by verifyingthat the motor circuit breakers are secured in the open position at least B-11Generic Issue 94once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS coldlegs is less than or equal to 2751F.Generally, it is preferable to include requirements for implementing the intentof an LCO as part of an LCO rather than to only define requirements, such assecuring motor circuit breakers in the open position, in a SurveillanceRequirement. Furthermore, the requirements for operable pumps could be statedin terms of requiring one pump to be operable rather in terms of "at leastone pump shall be operable" and then including a footnote requiring that, infact, no more than one pump shall be operable. The preferred alternativewould be an LCO which stated:One Safety Injection [and/or] charging pump shall be operable and allother Safety Injection [and/or] charging pumps shall be secured withtheir motor circuit breakers in the open position.The form of the above requirements for any given specification would be dependentupon which pumps are addressed by that specification, e.g., charging or injectionpumps or both.The surveillance requirements would be similar to that noted above with thefollowing substitution:...except the above required OPERABLE pump(s), shall be demonstrated tobe secured by verifying that the motor circuit breakers are in the openposition. ...Changes to plant TS should be proposed to incorporate one of the above methods,to ensure that pumps are not capable of initiating a mass addition to theRCS that exceeds the design basis assumptions for the LTOP system, for plantsthat do not currently include such requirements.The largest temperature increase in the RCS that could result in a challengeto the LTOP system is dependent upon the differential temperature between theRCS and the secondary system when starting a reactor coolant pump. Hence,this is also a matter important to safety when reactor coolant pumps are startedand the resulting increase in RCS temperature is in excess of the design basisassumption for the LTOP system to mitigate the resulting increase in RCSpressure. The manner in which this design basis assumption of the LTOP systemis reflected in TS has been the use of a footnote to the reactor coolant pumpoperability requirements to note that:A reactor coolant pump shall not be started with one or more of theRCS cold leg temperatures less than or equal to 2750F unless thesecondary water temperature of each steam generator is less than OFabove each of the RCS cold leg temperatures.The above footnote has been included in the TS for residual heat removalunder title of the Reactor Coolant System, Hot Shutdown.

B-12 Generic Issue 94Changes to plant TS should be proposed to incorporate the above method, to ensurethat the starting of RCS pumps are not capable of initiating a pressure transientthat exceeds the design basis assumptions for the LTOP system, for plants thatdo not currently have this requirement.

Generic Letter 90-06 -4 -For GI-94, the actions proposed by the NRC staff to improve the availabilityof the low-temperature overpressure protection (LTOP) system, as identified inSection 3 of Enclosure B, represent a new interpretation of existing requirementsfor some licensees and CP holders, and this request is considered a backfit inaccordance with NRC procedures. This backfit is a cost-justified safety enhance-ment. Therefore, an analysis of the type described in 10 CFR 50.109(a)(3) and50.109(c) was performed, and a determination was made that there will be asubstantial increase in overall protection of the public health and safety andthat the attendant costs are justified in view of this increased protection.The analysis and determination will be made available in the Public DocumentRoom with the minutes of the 167th and 168th meetings of the Committee toReview Generic Requirements.This request is covered by Office of Management and Budget Clearance Number3150-0011, which expires January 31, 1991. The estimated average burden hoursis 320 person-hours per licensee response, including assessment of the newrecommendations, searching data sources, gathering and analyzing the data, andpreparing the required reports. Comments on the accuracy of this estimate andsuggestions to reduce the burden may be directed to the Office of Managementand Budget, Room 3208, New Executive Office Building, Washington, D.C. 20503,and the U.S. Nuclear Regulatory Commission, Information and Records Management Branch,Office of Information Resources Management, Washington, D.C. 20555.Sincerely,OWigMat signed byJames C. PaStlowJames G. PartlowAssociate Director for ProjectsOffice of Nuclear Reactor Regulation

Technical Contact:

George A. Schwenk(301) 492-0878Enclosures:A. Staff Positions Resulting from Resolution of Generic Issue 70B. Staff Positions Resulting from Resolution of Generic Issue 94C. NUREG-1316, "Technical Findings and Regulatory Analysis Relatedto Generic Issue 70--Evaluation of Power-Operated Relief Valve and BlockValve Reliability in PWR Nuclear Power Plants"D. NUREG-1326, "Regulatory Analysis for the Resolution of GenericIssue 94, Additional Low-Temperature Overpressure Protection for Light-Water Reactors"9006200120DISTRIBUTION:Central Files C. Cheng L. Marsh R. Jones P. KadambiNRC PDR P04 Reading D. Pickett F. Hebdon H. SmithG. Holahan J. Partlow R. Scholl 0. JonesR. Baer, RES F. Gillespie W. Schwink W. Minners, RESK. Kniel, RES G. Schwenk R. Kirkwood, RES E. Throm, RES*Reviewed by (GL only)B. Caluro , Technical Edit~oj, on 1/26/90*SEE PREVIOUS CONCURRENCES: +OFC :PD4/LA _ :PD4/PM* :PD04/D :ADR4 (A)DR :ADPNAME :PNoonan tWqA0:DPickett:bj :FHebd :GHoIan :GHgan :JeaitlowDATE :01/24/90 :01/24/90 :01t3t/90 / / / /90 : 1*/I /90