ML20129E505
| ML20129E505 | |
| Person / Time | |
|---|---|
| Issue date: | 09/30/1996 |
| From: | Shapaker J NRC (Affiliation Not Assigned) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-96-06, GL-96-6, TAC-M96276, NUDOCS 9610010095 | |
| Download: ML20129E505 (1) | |
Text
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pn cs%q g* t UNITED STATES g j' t
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 001 i
I k * * * * * */ -September 30, 1996
[
i MEMORANDUM T0: Document Control Desk -
Document Management Branch Division of Information Support Services !
Office of Information Resources Management FROM: James W. Shapaker. Technical Assistant - J /
Events Assessment and Generic Communic s Bran DivisionofReactorProgramManagementp- '
Office of Nuclear Reactor Regulationt
SUBJECT:
DOCUMENTS ASSOCIATED WITH NRC GENERIC LETTER 96-06.
ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS (TAC No. M96276) 4 s
The Plant Systems Branch (SPLB) in the Divisison of Systems Safety and Analysis (DSSA) prt. pared the subject generic letter, which was issued on September 30. 1996, and given accession number 9609250096. There is material related to the subject generic letter that should be placed in the NRC Public Document Room and made available to the public. Therefore, by copy of this memorandum. I am providing the following documents to the NRC Public Document Room: (1) a cry of the published version of the subject generic letter.
(2) a copy of tie CRGR review package, and (3) a copy of the information paper (SECY-96-203) that was sent to the Commission.
I request that you provide me with the Nuclear Documents System accession number for this memorandum. This information may be provided by telephone (415-1151) or by e-mail (JWS). In addition, please modify the ap3ropriate NUDOCS entries to reflect the fact that the documents identified lerein are related to Generic Letter 96-06.
Attachments:
As stated 1
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9610010095 960930 PDR ORG NRRA PDR
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1 i UNITED STATES
. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 September 30, 1996 NRC GENERIC LETTER 96-06: ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS Addressees All holders of operating licenses for nuclear power reactors, except for those I licenses that have been amended to possession-only status.
Puroose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to (1) notify addressees about safety-significant issues that could affect con-tainment integrity and equipment operability during accident conditions, (2) request that all addressees submit certain information relative to the issues that have been identified and implement actions as appropriate to address these issuas, and (3) require that all addressees submit a written response to the NRC relative to implementation of the requested actions.
Backaround j As a result of recent NRC inspection activities, licensee notifications, and event reports, several safety-significant issues have been identified that I have generic implications and warrant action by the NRC to assure that these issues have been adequately addressed and resolved. In particular, the l following issues are of concern:
(1) Cooling water systems serving the containment air coolers may be exposed ,
to the hydrodynamic effects of waterhammer during either a loss-of- I coolant accident (LOCA) or a main steamline break (MSLB). These cooling l water systems were not designed to withstand the hydrodynamic effects of waterhammer and corrective actions may be needed to satisfy system design l and operability requirements.
(2) Cooling water systems serving the containment air coolers may experience two-phase flow conditions during postulated LOCA and MSLB scenarios. The heat removal assumptions for design-basis accident scenarios were based on single-phase flow conditions. Corrective actions may be needed to satisfy system design and operability requirements.
-- 9609250096 y' I
a~m',
.' GL 96-06 September 30, 1996
,.- Page 2 of 10 (3) Thermally induced overpressurization of isolated water-filled piping sections in containment could jeopardize the ability of accident-mitigating systems to perform their safety functions and could also lead to a breach of containment integrity via bypass leakage. Corrective actions may be needed to satisfy system operability requirements.
The sections that follow contain additional background information sbout each of these issues.
Waterhaemer On February 13. 1996, the Pacific Gas and Electric Company (PG&E. the licensee for Diablo Canyon Units 1 and 2), determined that component cooling water, which is circulated through the containment air coolers, could flash to steam in the cooler unit cooling coils during a design-basis LOCA with a concurrent loss of offsite power (LOOP) or with a delayed sequencing of equipment. This condition was reported to the NRC in Licensee Event Report (LER) 1-96-005, dated April 26, 1996.
The Diablo Canyon units have five containment air coolers in each containment, these are typically used during normal plant operation to prevent excessive containment temperatures. The containment air coolers are also automatically initiated engineered safety features that are relied upon to help maintain containment integrity by performing their heat removal function during postulated accident conditions. The air coolers in the Diablo Canyon units transfer heat from the containment to the respective unit's component cooling water system (a closed-loop system).
PG&E re)orted that, during a postulated design-basis LOCA with a concurrent LOOP. tie component cooling water pumps and the air cooler fans will tempo-rarily lose power (an expected condition). The component cooling water flow stops almost immediately, while the fans coast down over a period of minutes.
The first air cooler fan will restart on slow speed approximately 22 seconds after the LOOP and the component cooling water pumps will restart 4 to 8 seconds later. In this scenario the high-temperature containment atmo-sphere will be forced across the containment air cooler's cooling coils for up to 30 seconds with no forced component cooling water flow through the coolers.
PG&E determined that the stagnant component cooling water in the containment air coolers may boil and create a substantial steam volume in the component cooling water system. As the component cooling water pumps restart, the pumped liquid may rapidly condense this steam volume and produce a water-hammer. The hydrodynamic loads introduced by such a waterhammer event could be substantial, challenging the integrity and function of the containment air coolers and the associated component cooling water system, as well as posing a challenge to containment integrity. As corrective action. PG&E has installed a nitrogen pressurization system on the component cooling water head tank to increase the margin to boiling.
On June 20. 1996. Westinghouse Electric Corporation issued Nuclear Safety Advisory Letter NSAL-96-003. " Containment Fan Cooler Operation During a Design Basis Accident." to alert its customers to the potential safety issue that was
GL 96-06 September 30, 1996
- Page 3 of 10 identified by PG&E (Westinghouse is the reactor vendor for the Diablo Canyon units). In NSAL-96-003. Westinghouse recommended that licensees review their containment cooling systems to determine if their safety-related containment air coolers are susceptible to waterhammer.
On July 22. 1996, the Connecticut Yankee Atomic Power Company (CYAPC. the licensee for the Haddam Neck nuclear power plant) declared all four of the containment air coolers at the Haddam Neck plant inoperable and initiated a plant shutdown in accordance with Technical Specification requirements. The containment air coolers at the Haddam Neck )lant are the only components that are credited for post-accident containment leat removal, and station service water (an o>en-loop system) is the cooling medium for the containment air coolers. T1e containment air coolers were declared inoperable after CYAPC completed its review relative to Westinghouse NSAL-96-003. The licensee's analysis predicted hydrodynamic loads in the service water system from waterhammer that exceeded piping and support structural limits.
On August 12. 1996, the staff issued Information Notice (IN) 96-45 " Potential Comon-Mode Post-Accident Failure of Containment Coolers." to alert addressees to the potential failure mode of the containment air coolers and their associated cooling water systems. IN 96-45 discussed the information that was reported by PG&E and CYAPC relative to the Diablo Canyon and Haddam Neck plants, respectively, and attached a copy of Westinghouse letter NSAL-96-003.
Tw-Phase Flow in Safety-Related Piping and Components In July 1996, the NRC issued Inspection Report 50-213/96-201. "S)ecial i Ins)ection of Engineering and Licensing Activities at Haddam Nect-Connecticut i Yancee." Among other things, the report identified an issue relative to two-phase flow in the station service water system. The inspection team reviewed the service water system flow models, calculations, and operational data and found that some steam may be aroduced in the service water system as the service water flows through t1e containment air coolers during design-basis accident conditions. However, the licensee's service water system model and calculations only assumed single-phase flow conditions (liquid phase only) and did not consider two-phase flow conditions (both steam and liquid present).
The licensee is currently evaluating the system to determine whether or not corrective actions are needed.
On July 23, 1996, the Wisconsin Electric Power Company submitted information regarding two-phase flow in the service water system at the Point Beach nuclear plant during a design-basis LOCA. The licensee's preliminary evalua-tions concluded that after the cooling water is heated via heat transfer from the containment air coolers, some steam could be formed at the air cooler outlet throttle valves. This two-phase mixture (steam and water) would result in a higher frictional pressure drop in the service water return piping and
- would ultimately affect the service water flow and the heat removal capabili-
! ties of the containment air coolers. Steam formation due to low pressure and high temperature in the service water system could reduce the service water flow rates through the containment air coolers to values below those needed to
GL 96-06 September 30, 1996 ,
,, Page 4 of 10 ;
satisfy design-basis heat removal requirements. The licensee is completing more detailed analyses to determine if immediate corrective action is war-
. ranted.
On August 20, 1996, the Public Service Electric and Gas Company (the licensee
. for Salem 1 and 2) notified the NRC of a condition that is not bounded by the existing design basis for the Salem nuclear power plants (EN 30900). The ;
licensee reported that because the service water isolation valves for the nonsafety-related turbine loads do not start to close until approximately 30 seconds into the emergency loading sequence, the service water system may i not be able to supply sufficient flow for the containment cooling function ;
during accident conditions. The licensee determined that the initial heat transfer rates through the containment air coolers could result in additional 4 " flow restrictions in the air cooler tubes, further decreasing the flow of service water through the containment air coolers as a result of the higher
- frictional pressure drop caused by two-phase flow. At the time of the ,
licensee *s notification, the Salem units were shut down for refueling.
1 Overpressurization of Isolated Piping Sections
! On July 3. 1996. Duquesne Light Company (the licensee for Beaver Valley i Units 1 and 2) notified the NRC that during surveillance testing of a compo-l nent cooling water inlet valve to the RHR heat exchanger on Unit 1. the motor-o)erated butterfly valve located inside the containment would not open i ( EN 30833). The licensee found that pressure in the piping section between
! this valve and a closed manual butterfly valve located outside the containment .
. measured slightly higher than the system design 3ressure. After the pressure j in this isolated section of piping was relieved )y opening a drain valve, the !
remotely operated butterfly valve was opened without any trouble. The '
licensee concluded that pressure in the isolated section of piping increased when the trapped water was heated u) by increased ambient temperatures. The section of piping was isolated in tle spring when the unit was shut down and ambient temperatures were much lower than temperatures that existed in the summer after the plant was returned to power operation and ambient tempera-tures reached about 32 C f90 *F].
! On July 19, 1996, the Maine Yankee Atomic Power Company (MYAPC. licensee for the Maine Yankee nuclear )lant) notified the NRC of a condition that was outside the Jasis (EN 30769). The primary component cooling water (PCCW) plant at system design the Maine Yankee plant has a nonsafety-related subdivi-sion that serves the containment fan coolers (not needed for accident mitiga-tion) and a safety-related subdivision that serves ECCS equipment. The
', nonsafety-related subdivision of PCCW has a swing-check valve at the contain-ment inlet (supply) penetration, and an air-operated valve at the containment outlet (return) penetration. During a design-basis LOCA, the containment j isolation logic initiates closure of the air-operated outlet valve thereby '
stopping the flow of water. The licensee has determined that heat from the
- containment accident environment could cause the PCCW in the containment fan coolers between the inlet check valve and closed air-operated outlet valve to expand, rupturing this portion of the PCCW system. Water from the PCCW system 1 is then able to flow through the supply check valve for the containment fan l l
i
-~ GL 96-06 September 30, 1996
. Page 5 of 10 coolers and out the rupture, rendering the PCCW system inoperable and jeopar-dizing safety-related equipment that is cooled by the safety-related division of the PCCW system. Upon recognizing this postulated scenario, the licensee promptly shut down the Maine Yankee plant. To correct this, the licensee plans to install a pressure relief valve on each of the six containment fan .
cooler PCCW branch lines downstream of the supply check valves.
On August 20, 1996, the staff issued Information Notice (IN) 96-49, " Thermally Induced Pressurization of Nuclear Power Facility Piping," to alert addressees to the potential for safety-related piping to become overpressurized during accident conditions. IN 96-49 discusses the information reported by Duquesne Light Company and MYAPC relative to Beaver Valley Unit 1 and the Maine Yankee plant,respectively.
Discussion The issues discussed in this generic letter pertain to situations that may not be bounded by the applicable system design capabilities and for which correc-tive actions may be needed to satisfy equipment design and operability requirements. The sections that follow contain additional discussion about each of these issues.
Waterhamer At many plants. containment air coolers satisfy a significant safety function by removing heat from the containment and reducing post-accident containment pressure. The hydrodynamic loads imposed by waterhammer can be substantial, challenging the integrity and function of the containment air coolers and the associated cooling water system, as well as posing a challenge to containment integrity. Waterhammer in cooling water systems associated with nonsafety-related containment air coolers can also challenge containment integrity by creating a containment bypass flow path, and interfacing safety-related systems can be affected. During this accident scenario, the steam that is produced in the containment air coolers may accumulate in other parts of the cooling water system, restricting flow as well as causing waterhammer damage.
Plant vulnerability to the postulated waterhammer scenario depends on a number of factors, such as piping configuration, how long it takes for the flow of cooling water to stop, the coastdown rate of the fans in the containment fan coolers. the operating pressure and pressure decay rate of the cooling water system, how long it takes to establish forced cooling water flow, the contain-ment tenperature profile, and other site-specific parameters.
The postulated failure scenario is applicable to both LOCA and MSLB events that involve a loss of offsite power, a loss of cooling water flow to the containment air coolers (e.g., one train of cooling water inoperable), or the sequencing of equipment that can affect the containment cooling function.
Steam fonnation and waterhamer in cooling water systems associated with safety-related and nonsafety-related containment air coolers may not require a
, loss of offsite power for this scenario to be valid.
GL 96-06 September 30, 1996
. Page 6 of 10 Two-Phase Flow in Safety-Related Piping and Canponents Two-phase flow (i.e., both steam and liquid) in cooling water systems associ-ated with the containment air coolers can significantly interfere with the ability of the containment air coolers to remove heat under design-basis accident conditions, and can interfere with the cooling of other safety-related components. These cooling water systems were designed assuming single-phase flow conditions (i.e.. liquid only) and containment heat transfer analyses are based on this assumption. Two-phase flow is a much more complex situation to deal with analytically than single-phase flow and involves additional hydrodynamic loading considerations as well as flow, heat transfer, systems interaction and erosion considerations. Additionally, the steam that is formed during two-phase flow can accumulate in the cooling water system.
. restricting flow and resulting in a waterhammer as discussed above.
Overpressurization of Isolated Piping Because of its thermal expansion, water heated while it is trapped in isolated piping sections is capable of producing extremely high pressures. This
>henomenon is typically a design consideration. Piping design codes as far Jack as U.S.A. Standard (USAS) B31.1 (1967). have explicitly recognized the need to consider the effects of heating fluid that is trapped in an isolated section of piping. The potential for thermally induced expansion of fluid trapped in valve bonnets was one reason for issuing Generic Letter (GL) 95-07.
" Pressure Locking and Thermal Binding of Safety-Related Power-0perated Gate Valves." In addition, several information notices (ins) have been issued discussing the pressurization of water trapped in valve bonnets, including IN 95-14. " Susceptibility of Containment Sump Recirculation Gate Valves to i Pressure Locking." IN 95-18. " Potential Pressure-Locking of Safety-Related l Power-0perated Gate Valves." IN 95-30. " Susceptibility of LPCI and Core Spray Injection Valves to Pressure Locking." and IN 96 08, " Thermally Induced Pressure Locking of a HPCI Gate Valve." l The potential for systems to fail to perform their safety functions as a result of thermally induced overpressurization is dependent on many factors. 1 These factors include leak tightness of valve seats, bonnets, packing glands ,
and flange gaskets: piping and component material properties, location and I geometry; ambient and post-accident temperature response; pipe fracture mechanisms; heat transfer mechanisms: relief valves and their settings; and system isolation logic and setpoints. Engineering design and modification evaluations, which include systematic evaluation of heat input to systems and components with consideration of factors such as those just noted, can detect conditions which may influence system operability under normal operating, transient, and accident conditions.
Under the " single-failure concept." failure due to overpressurization does not preclude consideration of additional active and passive failures in the same !
and other systems in evaluating plant response to a postulated accident. If i relief valves are installed to prevent overpressure conditions, consideration I must be given to the effects of stuck-open relief valves and associated I environmental flooding and radiation hazards.
GL 96-06 September 30. 1996
. Page 7 of 10
\
Recuested Action (s)
Addressees are requested to determine:
(1) if containment air cooler cooling water systems are susceptible to either !
waterhammer or two-phase flow conditions during postulated accident !
conditions: >
(2) if piping systems that penetrate the containment are susceptible to thermal expansion of fluid so that overpressurization of piping could occur.
In addition to the individual addressee's postulated accident conditions.
these items should be reviewed with respect to the scenarios referenced in the generic letter.
With regard to waterhamer, addressees may find Volumes 1 and 2 of NUREG/CR-5220. " Diagnosis of Condensation-Induced Waterhammer," dated October 1988, informative and useful in evaluating potential waterhammer conditions.
If systems are found to be susceptible to the conditions discussed in this generic letter, addressees are expected to assess the operability of affected systems and take corrective action as appropriate in accordance with the requirements stated in 10 CFR Part 50 Appendix B and as required by the plant Technical Specifications. GL 91-18. "Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability." dated November 7, 1991, contains guidance on ,
the review of licensee operability determinations and licensee resolution of l degraded and nonconforming conditions. !
Reauested Information ,
Within 120 days of the date of this generic letter, addressees are requested to submit a written summary report stating actions taken in response to the requested actions noted above, conclusions that were reached relative to susceptibility for waterhammer and two-phase flow in the containment air cooler cooling water system and overpressurization of piping that penetrates containment, the basis for continued operability of affected systems and components as applicable, and corrective actions that were implemented or are planned to be implemented. If systems were found to be susceptible to the conditions that are discussed in this generic letter, identify the systems affected and describe the specific circumstances involved.
Reauired Resoonse Within 30 days of the date of this generic letter, addressees are required to submit a written response indicating: (1) whether or not the requested actions will be completed. (2) whether or not the requested information will be submitted and (3) whether or not the requested information will be sub-mitted within the requested time period. Addressees who choose not to
~
GL 96-06 September 30, 1996
. Page 8 of 10 complete the requested actions, or c. hoose not to submit the requested informa-tion, or are unable to satisfy the requested completion date, must describe in their response any alternative course of action that is pro)osed to be taken, including the basis for establishing the acceptability of t1e )roposed alternative course of action and the basis for continued opera)111ty of affected systems and components as applicable. i Address the required written reports to the U.S. Nuclear Regulatory Commis-sion, ATTN: Document Control Desk, Washington, D.C. 20555-0001, under oath or I affirmation, under the provisions of Section 182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). In addition, send a copy to the appropriate regional administrator. l l
Backfit Discussion i Title 10 of the Code of Federal Reculations (10 CFR) Part 50 (Appendix A) and plant licensing safety analyses require and/or commit that the addressees design safety-related components and systems to offer adequate assurance that those systems can perform their safety functions. Specifically,10 CFR Part 50, (Appendix A Criterion 38) specifies a " system to remove heat from the reactor containment. The safety function of this system is to rapidly reduce pressure and temperature in the containment following any loss-of-coolant accident and to maintain them at acceptably low levels." Addition-ally, Criterion 44 of Appendix A specifies a " system to transfer heat from
, structures, systems, and components important to safety. The system safety function shall be to transfer the combined heat load of these structures, .
systems and components under normal operating and accident conditions." The heat load values as defined in final safety analysis reports are based on single-phase flow assumptions for the containment air cooler cooling water systems. The potential for waterhammer and two-phase flow raises concerns :
that these systems will not meet their design-basis requirements as specified in 10 CFR Appendix A, Criteria 38 and 44. Further, 10 CFR Part 50 Appendix A, Criteria 1 and 4 specify that safety-related systems be designed to offer adequate assurance that those systems can perform their safety functions under accident conditions. Accordingly, licensees are required to ensure that the containment air coolers and their associated cooling water systems that may be .
l affected by waterhammer or by two-phase flow are capable of performing their required safety functions and that containment integrity will be maintained.
Licensees are also required either by their commitment to USAS B31.1 or the l American Society of Mechanical Engineers (ASME) Code for piping design or by l virtue of 10 CFR 50.55a, which endorses various editions of the ASME Boiler and Pressure Vessel Code, to comply with design criteria which specify that piping systems which have the potential to experience pressurization due to trapped fluid expansion shall either be designed to withstand the increased pressure or shall have provisions for relieving the excess pressure. The l
- potential for overpressurization raises concerns that these piping systems will not meet their design code criteria.
i The actions requested in this generic letter are considered compliance backfits under the provisions of 10 CFR 50.109 and existing NRC procedures to
GL 96-06 September 30, 1996 Page 9 of 10 :
ensure that containment integrity will be maintained and that safety-related components and piping systems are capable of performing their intended safety functions and satisfying their licensing-basis code criteria, respectively:
and that containment integrity and these safety-related piping systems and components will not be adversely affected by the occurrence cf waterhammer, two-phase flow, or thermal over)ressurization that may occur in safety-related and nonsafety-related systems tlat penetrate containment. In accurdarice with the provisions of 10 CFR 50.109 regarding compliance backfits, a full backfit l analysis was not performed for this aroposed action: but the staff performed a documented evaluation which stated tie objectives of and reasons for the requested actions and the basis for invoking the compliance exception. See also 10 CFR 50.54(f) . A copy of this evaluation will be placed in the NRC Public Document Room.
Federal Reaister Notification A notice of opportunity for public comment was not published in the Federal Reaister because of the urgent nature of the generic letter. However, comments on the actions requested and the technical issues addressed by this i generic letter may be sent to the U.S. Nuclear Regulatory Commission, ATTN Document Control Desk Washington, D.C. 20555-0001.
Pacerwork Reduction Act Statement )
This generic letter contains information collections that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were ap3 roved by the Office of Management and Budget, approval number 3150-0011, w11ch expires on July 31, 1997.
The public reporting burden for this collection of information is estimated to average 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the collection of information contained in the generic letter and on the following issues:
(1) Is the proposed collection of information necessary for the 3 roper performance of the functions of the NRC, including whether tie informa-tion will have practical utility?
- (2) Is the estimate of burden accurate?
(3) Is there a way to enhance the quality, utility, and clarity of the information to be collected?
(4) How can the burden of the collection of information be minimized, includ-ing the use of automated collection techniques?
I
,- GL 96-06 September 30, 1996 c
Page 10 of 10 Send comments on any aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Manage-ment Branch, T-6F33. U.S. Nuclear Regulatory Commission. Washington. D.C.
20555-0001, and to the Desk Officer. Office of Information and Regulatory Affairs. NE0B-10202 (3150-0011), Office of Management and Budget, Washington, D.C. 20503.
The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number.
If you have any questions about this matter, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
Thomas T. Martin, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts: Laura Dudes. NRR James Tatum. NRR (301) 415-2831 (301) 415-2805 Email: lad @nrc. gov Email: jet 1@nrc. gov John Fair (301) 415-2759 Email: jrf@nrc. gov Lead Project Manager: Beth Wetzel. NRR (301) 415-1355 Email: baw@nrc. gov
Attachment:
List of Recently Issued NRC Generic Letters
Attachment GL 96-06 c September 30, 1996
, Page 1 of 1 LIST OF RECENTLY ISSUED GENERIC LETTERS Generic Date of Letter Subject Issuance Issued To 96-05 PERIODIC VERIFICATION OF 09/18/96 ALL HOLDERS OF OLs DESIGN-BASIS CAPABILITY (EXCEPT THOSE LICENSES OF SAFETY-RELATED MOTOR- THAT HAVE BEEN AMENDED OPERATED VALVES TO POSSESSION-ONLY STATUS) OR cps FOR NPRs 96-04 BORAFLEX DEGRADATION IN 06/26/96 ALL HOLDERS OF OLs SPENT FUEL POOL STORAGE FOR NPRs RACKS 95-09, MONITORING AND TRAINING OF 04/05/96 ALL U.S. NUCLEAR SUPP. 1 SHIPPERS AND CARRIERS OF REGULATORY COMMISSION RADI0 ACTIVE MATERIALS LICENSEES 96-03 RELOCATION OF THE PRESSURE 01/31/96 ALL HOLDERS OF OLs TEMPERATURE LIMIT CURVES OR cps FOR NPRs AND LOW TEMPERATURE OVER-PRESSURE PROTECTION SYSTEM LIMITS 96-02 RECONSIDERATION OF NUCLEAR 01/31/96 ALL HOLDERS OF OLs POWER PLANT SECURITY OR cps FOR NPRs REQUIREMENTS ASSOCIATED WITH AN INTERNAL THREAT 89-10, CONSIDERATION OF VALVE 01/24/96 ALL HOLDERS OF OLs Supp. 7 MISPOSITIONING IN (EXCEPT THOSE LICENSES PRESSURIZED-WATER THAT HAVE BEEN AMENDED REACTORS TO A POSSESSION ONLY STATUS) OR cps FOR NPRs 96-01 TESTING OF SAFETY-RELATED 01/10/96 ALL HOLDERS OF OLs OR LOGIC CIRCUITS cps FOR NPRs 95-10 RELOCATION OF SELECTED 12/15/95 ALL HOLDERS OF OLs OR TECHNICAL SPECIFICATIONS cps FOR NPRs REQUIREMENTS RELATED TO INSTRUMENTATION OL - CPERATING LICENSE CP - CONSTRUCTION PERMIT NPR - NUCLEAR POWER REACTORS
GL 96-06 September 30, 1996
. Page 10 of 10 Send comments on any aspect of this collection of information. including suggestions for reducing this burden, to the Information and Records Manage-ment Branch, T-6F33. U.S. Nuclear Regulatory Commission, Washington, D.C.
20555-0001, and to the Desk Officer, Office of Information and Regulatory Affairs, NE0B-10202 (3150-0011), Office of Management and Budget, Washington, D.C. 20503.
The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number.
If you have any questions about this matter, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor
. Regulation (NRR) project manager.
Thomas T. Martin, Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts: Laura Dudes. NRR James Tatum, NRR (301) 415-2831 (301) 415-2805 Email: lad @nrc. gov Email: jet 1@nrc. gov John Fair 1 (301) 415-2759 Email: jrf@nrc. gov Lead Project Manager: Beth Wetzel. NRR (301) 415-1355 Email: baw@nrc. gov
Attachment:
List of Recently Issued NRC Generic Letters Tech Editor has reviewed and concurred on 08/30/96
- SEE PREVIOUS CONCURRENCES DOCUMENT NAME: 96-06.GL Ni.7C*i7.*Eo".U.$ YcN.E*,nInUnUl.'o,'. 'u~ .' E e*o',y ,
t 0FFICE TECH CONTS l OGC l C:PECB:DRPM l D:DRPy(//] l NAME LDudes* RHoefling* AChaffee* TJ4cf i _
JFair* ( '
JTatum* ,
DATE 08/21/96 09/05/96 09/06/96 / '//h/96 1 I
0FFJCIAL RECORD COPY
Attachment 2 CRGR REVIEW PACKAGE PROPOSED ACTION: Issue a generic letter on the potential waterhammer, two-phase flow and overpressurization events that have been reported by licensees. -
CATEGRY: 1 RESPONSE TO REQUIREMENTS FOR CONTENT OF PACKAGE SUBMITTED FOR CRGR REVIEW Question (1): The proposed generic requirement or staff position as it is proposed to be sent out to licensees. Where the objective or intended result of a )roposed generic requirement or staff position can be ac11eved by setting a readily quantifiable standard that has an unambiguous relationship to a readily measurable quantity and is enforceable, the proposed requirement should merely specify the objective or result to be attained, rather than prescribing to the li-censee how the objective or result is to be attained.
Response: Within 120 days of the date of this generic letter, address-ees are requested to submit a written summary report stat-ing, actions taken in response to the requested actions in the generic letter, conclusions that were reached relative to susceptibility for waterhammer and two-phase flow in the containment air cooler cooling water system and overpressur-ization of piping that penetrates containment, the basis for continued operability of affected systems and components as applicable, corrective actions that were implemented or are planned to be im)lemented. If systems were found to be susceptible to t1e conditions that are discussed in this generic letter, identify the systems affected and describe .
the specific circumstances involved. l 1
i Question (11): Draft staff 3 apers or other underlying staff documents supporting t1e requirements or staff positions. (A copy of all materials referenced in the document shall be made available upon request to the CRGR staff. Any Committee !
member may request CRGR staff to obtain a copy of any refer-ence material for his or her use.)
Response: On August 12. 1996, the NRC staff issued Information No-tice 96-45, " Potential Common-Mode Post-Accident Failure of Containment Coolers," in response to events at the Haddam Neck Nuclear Station and reports from the licensee for Diablo Canyon regarding the containment air cooler's suscep-tibility to component cooling water flashing in the cooler unit cooling coils under design-basis conditions.
2-On August 20, 1996, the NRC staff issued Information Notice 96-49. " Thermally Induced Pressurization of Nuclear Power Facility Piping." to alert addressees to a number of sce- !
narios reported by licensees that involved thermal expansion l of a fluid in closed piping that could lead to overpressuri- l zation of the piping.
These documents are provided as Attachments 3 and 4 of this memorandum. :
Question (iii): Each proposed requirement or staff position shall contain ;
the sponsoring office's position as to whether the proposal .
would increase requirements or staff positions, implement !
existing requirements or staff positions, or would relax or <
reduce existing requirements or staff positions.
Response: In accordance with NRC regulations 10 CFR Part 50 (Appendix A. criteria 38 and 44) and licensing commitments, and under the additional provisions of 10 CFR Part 50 (Appendix B.
Criterion XVI) licensees are expected to take actions to ensure that safety-related piping meets its licensing code criteria and that safety-related components are capable of performing their required safety functions under design-3 asis accident conditions.
All addressees are requested to review the safety-related piping and com]onents discussed in this generic letter for compliance wit 1 the current licensing basis requirements.
In addition, addressees are requested to submit the correc-tive action plans for any systems determined to be suscepti-ble to the issues described in this letter.
This generic letter implements existing NRC requirements.
Question (iv): The proposed method of implementation with the concurrence (and any comments) of OGC on the method proposed. The concurrence of affected program offices or an explanation of any nonconcurrence.
Response: The proposed method of implementation is the issuance of a generic letter. 0GC has no legal objection to this proposed generic letter.
Question (v): Regulatory analyses conforming to the directives and guid-ance of NUREG/BR-0058 and NUREG/CR-3568. (This does not apply for backfits that ensure compliance or ensure, define, or redefine adequate protection. In these cases a document-ed evaluation is required as discussed in IV.B.(ix).)
Response: This item is not applicable to this proposed staff action.
Question (vi): Identification of the category of reactor plants to which the generic requirement or staff position is to apply (that
3 3-is, whether it is to apply to new plants only, new OLs only.
OLs after a certain date, OLs before a certain date, all OLs, all plants under construction, all plants all water reactors all PWRs only, some vendor types, some vintage types such as BWR 6 and 4. jet pump and nonjet pump plants.
etc.).
Response: This proposed generic letter is applicable to all nuclear power plants.
Question (vii): For backfits other than compliance or adequate 3rotection backfits. a backfit analysis as defined in 10 C R 50.109.
The backfit analysis shall include, for each category of reactor )lants, an evaluation which demonstrates how the action s1ould be prioritized and scheduled in light of other ongoing regulatory activities. The backfit analysis shall document for consideration information available concerning any of the following factors as may be appro)riate and any <
other information relevant and material to t1e proposed action:
(a) Statement of the specific objectives that the proposed action is designed to achieve:
(b) General description of the activity that would be re-quired by the licensee or applicant in order to com-plete the action:
(c) . Potential change in the risk to the public from the accidental release of radioactive material:
(d) Potential impact on radiological exposure of facility employees and other onsite workers:
(e) Installation and continuing costs associated with the action, including the cost of facility downtime or the cost of construction delay:
(f) The potential safety impact of changes in plant or operational complexity, including the relationship of proposed and existing regulatory requirements and staff positions:
(g) The estimated resource burden on the NRC associated with the proposed action and the availability of re-sources:
(h) The potential impact of differences in facility type, design, or age on the relevancy and practicality of the proposed action:
(i) Whether the proposed action is interim or final, and if interim, the justification for imposing the proposed action on an interim basis:
(j) How the action should be prioritized and scheduled in light of other ongoing regulatory activities. The following information may be appropriate in this re-gard:
- 1. The proposed priority or schedule.
- 2. A summary of the current backlog of existing re-quirements awaiting implementation.
~
- 3. An assessment of whether implementation of existing requirements should be deferred as a result, and
- 4. Any other information that may be considered appro-priate with regard to priority, schedule, or cumu-lative impact. For example, could implementation be delayed pending public comment?
Response: This item is not applicable to this pro)osed staff action.
This action is considered a compliance Jackfit.
Question (viii): For each backfit analyzed pursuant to 10 CFR 50.109(a)(2)
(i.e., not adequate protection backfits and not compliance backfits), the )roposing Office Director's determination, together with t1e rational for the determination based on the consideration of paragraph (1) and (vii) above, that:
(a) There is a substantial increase in the overall protec-tion of public health and safety or the common defense and security to be derived from the proposal: and (b) The direct and indirect costs of implementation. for the facilities affected, are justified in view of this increased protection.
Response: This item is not applicable to this proposed staff action.
Question (ix): For adequate protection or compliance backfits evaluattd pursuant to 10 CFR 50.109(a)(4)
(a) a documented evaluation consisting of:
(1) the objectives of the modification (2) the reasons for the modification (3) the basis for invoking the compliance or adequate protection exemption.
(b) in addition, for actions that were immediately effec-tive (and therefore issued without prior CRGR review as
5-discussed in III.C) the evaluation shall document the safety significance and appropriateness of the action taken and (if applicable) consideration of how costs contributed to selecting the solution among various acceptable alternatives.
Response: Title 10 of the Code of Federal Regulations (10 CFR) Part 50 (Appendix A) and plant licensing safety analyses require and/or commit that the addressees design safety-related components and systems to offer adequate assurance that those systems can perform their safety functions. Specifi-cally,10 CFR Part 50. (Appendix A, Criterion 38) requires a, " system to remove heat from the reactor containment. The safety function of this system is to rapidly reduce pressure and temperature in the containment following any loss-of-coolant accident and to maintain them at acceptably low levels." Additionally, criterion 44 of Appendix A requires a, " system to transfer heat from structures, systems, and components important to safety. The system safety function shall be to transfer the combined heat load of these struc-tures, systems and com accident conditions." ponents under The heat loadnormal values operating as definedand in the licensees final safety analysis reports are based on single-phase flow assumptions for the containment air cooler cooling water systems. The potential for waterhammer and two-phase flow raises concerns that these systems will not meet their design-basis requirements as discussed in 10 CFR Appendix A, criterion 38 and 44.
The design code criteria as defined in each licensees final safety analysis report, has provisions which require piping systems that have the potential to experience pressurization due to trapped fluid expansion be designed to withstand the increased pressure or provisions shall be made to relieve the excess pressure. The potential for overpressurization raises concerns that these piping systems will not meet their design code criteria.
In accordance with those regulations and licensing commit-ments, and under the additional provisions required by 10 CFR (Appendix A Criteria 1 and 4), which requires that safety-related systems be designed to offer adequate assur-ance that those systems can perform their safety functions under accident conditions, licensees are required to ensure that the containment air coolers and their associated cool-ing water systems that me be affected by waterhammer or by two-phase flow are capable of performing their required safety functions and that containment integrity will be maintained.
The actions requested in this generic letter are considered compliance backfits under the provisions of 10 CFR 50.109 and existing NRC procedures to ensure that containment
integrity will be maintained and that safety-related compo-nents and piping systems are capable of performing their intended safety functions and satisfying their licensing-basis code criteria, respectively: and that containment ;
integrity and these safety-related pi aing systems and compo-nents will not be adversely affected )y the occurrence of waterhammer, two-phase flow, thermal overpressurizatio'n that may occur in safety-related and non-safety-related systems that penetrate containment. In accordance with the provi-sions of 10 CFR 50.109 regarding compliance backfits, a full backfit analysis was not performed for this proposed action; but the staff performed a documented evaluation which stated the objectives of and reasons for the requested actions and the basis for invoking the compliance exception.
In accordance with those regulations and licensing commit-ments and under the additional provisions of 10 CFR Part 50 (Appendix B Criterion XVI), licensees are expected to take actions to ensure that:
(a) the containment air coolers and their associated cool-ing water systems are capable of performing their re-quired safety function, (b) any safety-related piping subject to thermally induced overpressurization will still maintain its structural integrity as defined by the construction code of re-cord.
Question (x): For each evaluation conducted for proposed relaxations or decreases in current requirements or staff positions, the proposing Office Director's determination, together with the rationale for the determination based on the considerations i or paragraphs (i) through (vii) above, that:
l (a) The public health and safety and the common defense and security would be adequately protected if the proposed l 2
reduction in requirements or positions were imple- !
mented, and (b) The cost savings attributed to the action would be sub-stantial enough to justify taking the action.
Response: This item is not applicable to this proposed staff action.
Question (xi): For each request for information under 10 CFR 50.54(f)
(which is not subject to exception as discussed in III.A) an evaluation that includes at least the following elements:
(a) A problem statement that describes the need for the information in terms of potential safety benefit.
4 ,
7 (b) The licensee actions required and the cost to develop a response to the information request.
(c) An anticipated schedule for NRC use of the information.
(d) A statement affirming that the request does not im)ose new requirements on the licensee, other than for tle requested information.
Response: Through 10 CFR 50.54(f). in this proposed generic letter the staff requires that addressees and permit holders notify the staff whether or not the actions requested in the generic letter will be implemented. The staff also requires ad-dressees to reply in writing to the request to submit a summary description of those systems found to be susceptible to the problems described in the letter, the corrective actions to be taken in order to meet the current licensing .
code criteria for those systems, and a schedule of implemen- I tation of these actions. The safety benefit of the staff's receiving the information is that the staff will be able to verify that addressees are in compliance with their current licensing basis and to compare the responses in order to further analyze those systems that could represent a poten-tial safety hazard to nuclear power plant o)eration. The staff will use this information to establisi appropriate inspection schedules for those addressees considered to need a more detailed staff review. The information requirement ,
does not impose new requirements on the licensee, other than
, for the requirements to prepare the information identified in the generic letter. ;
Question (xii): An assessment of how the 3roposed action relates to the Commission's Safety Goal Jolicy Statement.
Response: The staff considers the proposed generic letter to be con-sistent with the Commission's Safety Goal Policy Statement because it is a compliance backfit and not a " safety en-hancement." ;
1
AC6 t
l
.,. t POLICY ISSUE 4
(Information) -
1 l
September 20, 1996 SECY-96-203 l FAR: The Commissioners
?
faQM: James M. Taylor Executive Director for Operations
SUBJECT:
PROPOSED GENERIC LETTER TITLED " ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS ACCIDENT CONDITIONS" PURPOSE:
The purpose of this information paper is to inform the Commission, in accordance with the guidance in the December 20, 1991, memorandum from Samuel J. Chilk to James M. Taylor regarding SECY-91-172, " Regulatory Impact Survey Report-Final," of the staff's intent to issue the subject generic letter. The generic letter requests addressees to review cooling water systems that serve the containment air coolers (CACs) and determine if they are susceptible to waterhammer or to two-phase flow development, or both during postulated accident conditions. Additionally, the generic letter requests addressees to determine if piping systems which penetrate the containment are susceptible to ,
thermal expansion of fluid so that overpressurization may occur. A copy of l the proposed generic letter is attached. ]
DISCUSSION:
As a result of recent NRC inspection activities, licensee notifications, and event reports, several safety-significant issues have been identified that have generic implications and warrant action by the NRC to assure that these issues have been adequately addressed and resolved. The generic letter alerts :
addressees to several regulatory concerns pertaining to two-phase flow and waterhammer in the CAC cooling water systems, and thermally induced overpressurization of isolated sections of piping'inside containment. l l
I
Contact:
George Hubbard, NRR NOTE: TO BE MADE PUBLICLY AVAILABLE IN (301) 415-2870 5 WORKING DAYS FROM THE DATE OF THIS PAPER L _Shhf
The Commissioners -
2-The potential for waterhammer in the cooling water system of the containment air coolers was addressed in the Diablo Canyon Licensee Event Report (LER,1-96-005) dated April 26, 1996. After reviewing the Diablo Canyon event, Westinghouse Electric Corporation issued a Nuclear Safety Advisory Letter recommending that its customers review the CAC cooling water systems for susceptibility to a postulated waterhammer event. While performing this recommended review, Connecticut Yankee Atomic Power Compmy ( the licensee for the Haddam Neck Nuclear Station) determined that there was a potential for the postulated waterhammer to cause hydrodynamic loads that exceed the current design-basis loads in the cooling water piping. Therefore, the licensee shut down the plant. The hydrodynamic loads imposed by waterhamer could potentially challenge the integrity and function of the CACs as well as challenge the integrity of the containment itself via bypass leakage. Steam produced in the CACs may accumulate in other parts of the cooling water system restricting flow and establishing the necessary conditions for structural damage as a result of waterhamer. A plant's vulnerability to waterhamer is dependent on many factors, such as piping configurations, sequencing of equipment, and the containment temperature profile. The staff issued infonnation notice (IN) 96-45, " Potential Comon-Mode Post-Accident Failure of Containment Coolers," on August 12, 1996 to alert the industry to this problem.
The potential for two-phase flow in the CAC cooling water system is also addressed in the generic letter. As a result of inspection activities at the Haddam Neck facility, the NRC became aware of certain loss-of-coolant- .
accident conditions which could lead to the development of two-phase flow in the CAC cooling water piping. This condition was also found to exist at the Point Beach and Salem nuclear power plants. Two-phase flow in the CAC cooling water system affects the ability of the CAC to remove its design-basis heat loads, could interfere with the cooling of other safety-related components and
, could potentially cause waterhamerJ transients to occur. _
l The third and final issue described in the generic letter is the potential overpressurization of isolated piping sections within the containment.
Overpressurization can occur when fluid is trapped between two valves and subsequently heated by the increased temperatures in containment during
- accident conditions. This condition has been identified at several operating nuclear plants, including Beaver Valley, Maine Yankee and Vermont Yankee.
The overpressurization of piping from fluid expansion has been clearly addressed by U.S.A. Standard B31.1 piping design criteria and the American Society of Mechanical Engineers Code. Both codes have provisions which specify that systems with the potential for this event must be designed to withstand the increased pressure or must contain a path to relieve excess pressure. The staff issued IN 96-49, " Thermally Induced Pressurization of l Nuclear Power Facility Piping," on August 20, 1996, to alert the industry of 1
this problem.
The staff has had several meetings and conference calls with licensees and
! regional inspectors as these issues were identified at the individual plants.
i Currently, the staff is dealing with these issues on a case-by-case basis; the generic communication will disseminate these staff safety concerns to all i
-y - -
I , .
1:
e The Commissioners i operating reactors and will provide for the identification of these issues and corrective actions in a broad manner. The attached proposed generic letter i requests that licensees review their systems for susceptibility to these i
problems to ensure the containment air coolers are capable of performing their
- safety function and that they are in compliance with the piping design criteria. The generic letter requests that licensees submit a written summary i report stating actions taken in response to the issues identified in the i
i generic istter, conclusions reached regarding individual plant susceptibility to these problems, the basis for continued operability, as applicable, and
- corrective actions taken. .
4 A notice of opportunity for public comment on the proposed generic letter will
- not be published in the Federal Register because of the urgent nature of the 4 generic communication.
The generic letter is considered urgent for the following reasons: (1) the identified problems exist at several operating plants and hardware modifications have been made or are planned; (2) the staff believes that all plants are potentially susceptible to the problems identified in the generic
! letter and (3) two operating plants have already shut down due to the issues j identified in this generic letter (Maine Yankee, Haddam Neck).
! The Office of the General Counsel has reviewed the proposed generic letter and i has no legal objection to its issuance.
i
! Thc proposed generic letter was transmitted to the Committee to Review Generic
! Requirements (CRGR) by a memorandum from Frank J. Miraglia to Edward L. Jordan l on September 6, 1996. The CRGR was briefed by the staff on September 11, i
1996. Following the briefing and incorporation of CRGR comments, the CRGR endorsed the issuance of the pro)osed urgent generic letter. The package reviewed by the CRGR contained tie staff's responses to the CRGR Charter l questions; this package will be placed in the NRC's Public Document Room.
i The staff intends to issue this generic letter approximately five working days j after the date of this information paper.
I 4 f
i a s . Tay r i ecutive Director j for Operations
Attachment:
Proposed Generic Letter 96-XX 1 DISTRIBUTION:
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