Information Notice 1987-50, Potential LOCA at High- and Low-Pressure Interfaces From Fire Damage

Potential LOCA at High- and Low-Pressure Interfaces From Fire Damage

ML070180060
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, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill
Issue date:	10/09/1987
Revision:	0
From:	Rossi C E Office of Nuclear Reactor Regulation
To:
References
IN-87-050, NUDOCS 8710060172
Download: ML070180060 (4)

CCTKIC No.IN 87-50UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555October 9, 1987NRC INFORMATION NOTICE NO. 87-50: POTENTIAL LOCA AT HIGH- AND LOW-PRESSUREINTERFACES FROM FIRE DAMAGE

Addressees

All nuclear power reactor facilities holding an operating license or a con-struction permit.

Purpose

This information notice is provided to alert recipients to a potentiallysignificant safety problem pertaining to the possible initiation of aloss-of-coolant accident (LOCA) as a result of fire damage in the controlroom or the cable spreading room. If the postulated fire causes a hotshort which opens a high pressure to low-pressure system interface isolationvalve, exposure of the low-pressure system to pressures in excess of itsdesign pressure could result in a LOCA. It is expected that recipients will.review the information for applicability to their facilities and consideractions, if appropriate, to preclude a similar problem. However, suggestionscontained in this notice do not constitute NRC requirements; therefore, nospecific action or written response is required.Background:The requirements of 10 CFR Part 50, Appendix R, "Fire Protection Program forNuclear Power Facilities Operating Prior to January 1, 1979," are applicableto all licensed nuclear power reactor facilities that were operating beforeJanuary 1, 1979. Facilities that were licensed after that date either commit-ted to comply with the requirements of Appendix R or were reviewed forconformance with the guidelines of the Standard Review Plan (NUREG-0800),Section 9.5.1, "Fire Protection Program," which incorporates the requirementsof Appendix R as guidelines. Thus, the same criteria have been used on allnuclear power reactor facilities. In either case, they are simply referredto as the criteria of Appendix R for the purpose of this information notice.Appendix R states, in part, that where adequate fire protection of safe shut-down systems cannot be maintained, an alternative method of safely shuttingdown the plant shall be provided. For most plants, an alternate shutdownmethod is required in the event of a postulated fire in the control room orthe cable spreading room. Appendix R also states that for these areas, "...thefission product boundary integrity shall not be affected, i.e., there shall be8706017 4Z ,- T EA ýc e.9~

IN 87-50October 9, 1987 no...rupture of any primary coolant boundary." Thus, for those low-pressuresystems that connect to the reactor coolant system (a high-pressure system),at least one isolation valve must remain closed despite any damage that maybe caused by fire. A fire could occur in the panel or cables that controlthe isolation valves causing hot shorts that may result in opening the valvesat the high/low-pressure interface. Since the low-pressure system could bedesigned for pressures as low as 200 to 400 psi, the high pressure from thereactor coolant system (approximately '1000 to 1200 psi for BWRs and 2000 to2200 psi for PWRs) could result in failure of the low-pressure piping. In manyinstances, the valves at the high-pressure to low-pressure interface are notdesigned to close against full reactor coolant system pressure and flow con-ditions. Thus, spurious valve opening could result in a LOCA that cannot beisolated, even if control of the valve can be reestablished.

Description of Circumstances

During a fire protection re-review at Washington Public Power Supply System'sWashington Nuclear Project Number 2 (WNP-2), the licensee discovered thatshould a fire occur in the control room, power would have to be removed fromthe valve motor operators in the residual heat removal (RHR) system suctionand discharge lines to prevent inadvertent valve operations resulting frompossible fire damage to the circuits. If the damage occurred before removingpower to the valve motor operators, the valves could be spuriously opened,resulting in overpressurization of the RHR piping that could lead to a LOCAthat could not be isolated.In discussions with the WNP-2 personnel, the NRC staff became aware of a bypassline around the check valve in the discharge line that had a motor-operatedisolation valve in the line. This bypass line is used to warm up the RHRsystem discharge line by backflow from the reactor before initiating residualheat removal to prevent thermal shocking of the reactor vessel nozzle safe end.Becau-.e of this bypass line around the check valve, credit for the check valvein preventing a LOCA at the high- and low-pressure interface can no longer begiven.The licensee intends to remove the power to this motor operator during normalpower operations. Since this valveis used only for prewarming the RHR lineduring a normal shutdown, removing power during normal power operations shouldnot adversely impact safe plant operations.In order to determine if other plants have piping designs similar to that ofWNP-2, the final safety analysis reports of nine other BWRs were reviewed bythe staff. These included BWR-4, BWR-5, and BWR-6 designs. Of these nineplants, six (Clinton Power Station; Hope Creek Nuclear Station; LimerickGenerating Station; Nine Mile Point Nuclear Station, Unit 2; Perry NuclearPower Plant; and Susquehanna Steam Electric Station) have a piping configu-ration simila- to that of WNP-2. One plant (Monticello Nuclear GeneratingPlant) has a design similar to WNP-2 but has two normally closed, locallyoperated manual valves in the bypass line; therefore, this problem does notappear to apply to this plant. The two remaining plants (Grand Gulf NuclearStation and River Bend Station) do not have bypass lines around the check valv IN 87-50October 9, 1987 low-pressure inter-The potential for creating a LOCA from a similar high- andface may also be applicable to PWRs.No specific action or written response is required by this information notice.If you have any questions about this matter, please contact the RegionalAdministrator of the appropriate regional office or this office.Charles E. Rossi, DirectorDivision of Operational Events AssessmentOffice of Nuclear Reactor Regulation

Technical Contact:

John N. Ridgely, NRR(301) 492-4742

Attachment:

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