ML20199C680

Provides Encl Documents Associated W/Nrc GL-97-04, Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling & Containment Heat Removal Pumps

ML20199C680
Person / Time
Issue date:	11/18/1997
From:	Shapaker J NRC (Affiliation Not Assigned)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-97-04, GL-97-4, TAC-M96537, NUDOCS 9711200094
Download: ML20199C680 (18)

Text

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UNITED STATES g-g g

NUCLEAR REGULATORY COMMIS810N WASHINGTON. D.C. 30645 4001

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November 18 1997 MEMORANDUM T0:

Document Control Desk Information and Records Management Branch Information Mana ement Division Office of the Ch ef Inprmation Off er FROM:

James W. Shapaker h b fr/ 4

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Events Assessment.l eneric Communicat'Gns and Special Inspections Branch Division of Reactdb Program Management Office of Nuclear Reactor Regulation'

SUBJECT:

DDCUMENTS ASSOCIATED WITH NRC GENERIC LETTER 97-04.

ASSURANCE OF SUFFICIENT NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL PUMPS (TAC NO. M96537)

The Containment Systems and Severe Accident Branch Branch (SCSB) in the Divisison of Systems Safety and Analysis-(DSSA) prepared the subject generic letter. which was issued on October 7.1997, and given accession number 9710060324. There is material related to the subject generic letter that should be ) laced in the NRC Public Document Room and made available to the public. Tierefore. by copy of. this memorandum. I am providing the following documents to the NRC Public Document Room:

(1) a copy of the published version of the subject generic letter. (2) a copy of the information paper (SECY-97-217) that was sent to the Commission. (3) a co)y of each letter eceived in response to the notice of o)ortunity for pu)lic comment on the r

)roposed generic letter that was publisled in the Federal Register on r bruary 20. 1997. (4) a copy of the summary and resolution of public comments e

that were received, and (5) a copy of the CRGR review package.

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 97-04.

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e UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-C001 October 7,1997 NRC GENERIC LETTER 97-04: ASSURANCE OF SUFFICIENT NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL PUMPS Addressees All holders of operating licenses for nuclear power plants, except those who have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.

PJLrpan The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter (GL) to request that addressees submit information necessary to confirm the adequacy of the net positive suction head (NPSH) available for emergency core cooling (including core spray and decay heat removal) and containment heat removal pumps.

Ilackaround As a result of recent inspection activities, licensee notifications, and licensee event reports (LER), the NRC has identified a safety significant issue that has generic implications and warrants action by the NRC to ensure that the issue is adequately addressed anri resolved.

The issue is that the NPSH available for ememency core cooling system (ECCS) (including core spray and decay heat removal) cnd containment heat removal pumps may not be adequate under all design-basis accident scenarios, in some cases, this inadequacy may be a result of changes in plant configuration, operating procedures, environmental conditions, or other operating parameters over the life of the plant.

In other cases, a plant's NPSH analysis may not bound all postulated events for a sufficient time, or assumptions used in the analysis may be non-conservative or inconsistent with assumptions and methodologies traditionally considered acceptable by the staff. For example, some licensees have recently discovered that they must take new or additional credit for containment overpressure to meet the NPSH requirements of the emergency core cooling system and containment heat removal pumps. In the examples the NRC staff is familiar with, the need for crediting this overpressure in NPSH analyses has arisen because of changes in plant configuration and operating conditions, and/or errors in prior NPSH calculations. As a result, the overpressure being credited by licensees may be inconsistent with the plant's respective licensing basis.

-9710060324' I [

GL 97-04 October 7,1997

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Page 2 of 10 Current NPSH analyses (including any corresponding containment pressure analyses) may not be available to the staff in docketed material (such as final safety analysis reports) because some licensees nave changed their analyses. Consequently, this generic letter requests that addressees provide current information regarding the NPSH analyses for emergency core cooling and containment heat removal pumps. This generic letter applies only to ECCS and containment heat removal pumps that meet the following cnteria:

(1) pumps that take suction from the containment sump or suppression pool following a casign-basis loss-of-coolant accident (LOCA) or secondary line break, or (2) pumps used in " piggyback" operation that are necessary for recirculation cooling of the reactor core ano containment (that is, pumps that are supplied by pumps which take suction directly from the sump or suppression pool).

New NPSH analyses are neither requested nor required to be performed to respond to this information request. However, new NPSH analyses may be warranted if an addressee determines that changes in plant design or procedures have occurred which may have reduced the available NPSH. In such cases, each affected addressee must take appropriate corrective action to restore its facility to compliance, in accordance with the requirements stated in Appendix B to 10 CFR Part 60.

The following is a sample of the NRC staff's recent findings conceming the NPSH issues addressed by this genoric letter, Haddam Neck In 1986 and 1995, the licensee identified conditions for which the NPSH available for residual heat removal (RHR) pumps may be insufficient when the pumps are operating in the emergency core cooling mode. In 1986, the licensee determined that the only extant NPSH analysis, which was performed in 1979 as part of the Systematic Evaluation Program, did not properly account for hydraulic losses in suction piping. As a result, that analysis erroneously indicated that containment overpressure was not needed to satisfy NPSH requirements for the pumps in the recirculation mode of operation. A subsequent analysis showed that the licensee needed to take credit for 41.36 kPa (6 psig) of containment overpressure. In another analysis conducted in 1995 using increased service water temperature, the licensee found that additional containment overpressure was necessary to meet NPSH requirements for the same pumps. This additional overpressure constituted a significant fraction of the peak calculated containment accident pressure.

GL 97-04 October 7,1997 Page 3 of 10 On August 30,1996, the licensee reported in LER 96-016 that calculations recently performed to determine the NPSH available fcr the RHR pumps may have been in error for the attemate.nort term recirculation flow path, because of insufficient containment ove :,ure for a period of pump operation. The licensee attributed this error to its failure to fully analyze the containment pressure and sump temperature responses under design.

basis accident conditions.

Meine Yankee in July and August 1996, an NRC Independent Safety Assessment Team (lSAT) conducted an inspection to determine if Maine Yankee was operating in conformance with its design and licensing bases. During that inspection, the ISAT identified potential weaknesses in the NPSH analysis conducted by the licensee for the containment spray pumps. These potential weaknesses included concems regarding the validity of the containment sump temperature analysis, incorrect calculation of bounding pump suction head losses, and use of a hot-fluid correction factor to reduce NPSH requirements.

The licensee's calculation of record, performed in 1995 for a power level of 2700 thermal megawatts (MWt) and which does not include the hot fluid correction factor, indicates that the available NPSH for the containment spray pumps would be below the required NPSH for the first 5 minutes after pump suction is switched from the refueling water storage tank to the recirculation sump. When the licensee repeated the analysis using the hot fluid correction factor (the use of which the ISAT viewed as a non-conservative assumption as implemented by Maine Yankee), the available NPSH was only slightly greater than the required NPSH for the same 5 minute period. For the remainder of the transient, the licensee's analysis showed that NPSH available to the containment spray pumps would exceed the amount required. As a basis for the contention that the containment spray pumps were operable despite the 5-minute period with available NPSH below the required NPSH, the licensee cited recent pump tests showing that the pumps could operate for a 15-r.iinute period with NPSH below the required value without damage to the hydraulic performance or mechanical integr y of the pumps.

The licensee performed another analysis for a power level of 2440 MWt, which showed that adequate NPSH margin would be available for the containment spray pumps in the recir-culation mode of operation. This analysis did not include use of the hot-fluid correction factor. The ISAT concluded that it was appropriate to consider the containment spray pumps operable at a power level of 2440 MWt.

Pilgrim As indicated in the NRC safety evaluation for licensing of the Pilgrim plant, and in documents referenced by that evaluation, containment overpressure was not necessary to satisfy RHR and core spray pump NPSH requirements at the time of licensing. When the plant was modified in 1984, the licensee's safety evaluation related to the modification stated that the available NPSH was determined assuming (1) maximum debris loading conditions on the sump strainers for the RHR and core spray pumps and (2) no credit for containment

GL 97-04 October 7,1997 Page 4 of 10 overpressure. The licensee reaffirmed this assumption on April 14,1994, in its response to NRC Bulletin 93-02,

• Debris Plugging of Emergency Core Cooling Suction Strainers," dated March 23,1993, stating that the NPSH available to the residual heat removal and core spray pumps was analyzed assuming no overpressure condition in the torus.

However, in an analysis conducted by the licensee in 1995 in support of a proposal to raise the design seawater injection temperature to 75'F, credit was needed and taken for contain-ment overpressure. At the time of this analysis, the licensee also indicated that the assumption of no overpressure in the torus, stated in its response to Bulletin 9342, vns incorrect. This example illustrates that the potential exists that other licensees may have made modifications to their plants that could be inconsistent with the plant's licensing basis, and could reduce the NPSH available to the ECCS pumps.

Crystal River, Unit 3 in July 1996, an NRC inspection team conducted sin integrated Performance Assessment of Crystal River, Unit 3. As part of that assessment, the team reviewed the licensee's calculation which established the minimum post-LOCA reactor building water level required to ensure that adequate NPSH would be available for the reactor building spray pumps. When the team compared this level with the minimum predicted level, they found that for one of the pumps, there was only a slight difference between the available water level and the level required to ensure adequate NPSH dunng the post accident recirculation phase of pump operation.

The taam found that the licensee used non-conservative assumptions in calculating the available NPSH for the spray pump. For example, the licensee failed to account for uncertainty in data regarding the required NPSH, as well as for uncertainties associated with with the hydraulic resistance of check valves in the spray lines. In addition, the licensee used a hot fluid correction factor to reduce the required NPSH without considering the effects of non-condensable gases in the pumped fluid. Conservative assumptions included in the licensee's calculation were those detailed in Regu!atory Guide (RG) 1.1,

• Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps,"

dated November 2,.1970 (originally Safety Guide 1), regarding the use of maximum reactor building fluid temperature and lack of credit for containment overpressure.

The team concluded that the non-conservative assumptions used in the licensee's NPSH calculation raise questions conceming the cavitation free operation of reactor building spray pump 1B during the recirculation phase o operation. However, the team also concluded that this issue did not constitute an immediate safety concem since the licensee's calculations conservatively assumed no credit for containment overpressure and used the maximum expected reactor building water temperature.

Dresden By letter dated January 13,1997, the licensee for Dresden submitted a license amendment request for approval of 13 KPa (2 psig) of containment overpressure for the first 10 minutes

GL 97-04 October 7,1997 Page 5 of 10 following a design-basis LOCA. This overprescure is necessary to compensate for an NPSH deficiency for the low pressure coolant injection (LPCI) and core spray pumps. The licensee identified the need for overpressure after discovering that an incormet value for the ECCS suction strainer head loss had been used in the design basis NPSH calculation. As part of a design-basis review, the licensee determined that the actual head loss across the suction strainers was 1.8 m (5.8 feet) for clean strainers, rather than the 0.30 m (1 foot) head loss assumed in Dresden's original design basis as documented in the final safaty analysis report and vendor draw'ngs.

Because the licensee could not determine w44 certainty if overpressure was part of the original Dresden licensing basis, the licenst : 'oncluded that the use of overpressure constituted an unreviewed safety question and therefore requested staff approval to credit overpressure. In a licenso amendment dated January 28,1997, the staff approved the requested use of 13 kPa (2 psig) of containment overpressure, h a subsequent license amendment issued on April 30,1997, the staff approved tha use of a maximum of 65 kPa (9.5 psig) of containment overpressure for NPSH, for the first 240 seconds following a design basis LOCA. The need for this greater amount of overpressure arose prirnarily because of a higher calculated suppression pool temperature than that used in the analysis to support 13 kPa (2 psig) of overpressure.

Monticello in a report submitted to the NRC on Apnl 15,1997, pursuant to 10 CFR 50.72, the licensee for Monticello reported that the NPSH available to its core spray pumps may not meet the required NPSH under all accident conditions. The licensee discovered this possibility during a review of ECCS pump NPSH requirements, when a higher head loss than had previously been assumed for the ECOs suction strainers was calculated. During discuss;ons with the licensee, the staff leamed that the head loss across the suction strainers is approximately 3.57 m (11.7 feet) per 38.000 liters / minute (10,000 gpm), rather than the 0.3048 m (1 foot) per 38,000 liters / minute (10,000 gpm) assumed in the original design-basis analysis.

The licensee determined that for a recirculation line break with a single failure of the LPCI loop select logic, and with credit for containment overpressure, the core spray pumps would have an NPSH deficit and the LPCI pumps would have approximately 0.15 rn (0.5 feet) of margin in NPSH. Following discovery of the NPSH condition, the licensou conducted an operability evaluation of the LPCI anc' core spray pumps, and made this evaluation available to the staff for review. Subsequently, on May 9,1997, the licensee for Monticello commenced a voluntary shutdown of the plant because of the possible NPSH deficit for the ECCS pumps that would occur as a result of postulated clogging of the ECCS suction strainers under design basis LOCA conditions.

Related Genene Communications On October 22,1996, the staff issued information Notice (IN) 96-55, " inadequate Net Positive Suction Head of Emergency Core Cooling and Containment Heat Removal Pumps Under Design Basis Accident Conditions," to alert addressees to recent discoveries by

GL 97-04 October 7,1997 Page 6 of 10 licensees of possible scenarios for which the NPSH available for ECCS and conte.inment heat removal pumps, is insufficient. Earlier ins describing similar events include IN 87-63,

" inadequate Nei Positive Suction Head in Low Pressure Safety Systems," dated December 9,1987, and IN 88 74, "Potentially inadequate Perfonnance of ECCS in PWRs During Recirculation Operation Following a LOCA," dated September 14,1988.

DISCUSSION it is important that the emergency core cooling (including core spray and decay heat removal) and containment spray system pumps have adequate NPSH available to ensure that the systems can reliably perform their intended functions under all design-basis LOCA conditions. Inadequate NPSH could cause voiding in the pumped fluid, resulting in pump

< m /itation. While some ECCS and containment hett removal pumps can oper* 'cr rela-uvely snort periods of time while cavitating, prolonged operation of any pump under cavita-tion conditions can cause pump damage with potential common-mode failure of the pumps.

Such common mode failure would result in the inability of the ECCS to provide adequate long-term core cooling and/or the inability of the containment spray system to maintain the containment pressure and temperature below design limits.

This generic letter addresses situations in 'vhich the NPSH available to the ECCS and containment heat removal pumps may be inadequate as a result of changing plant conditions and/or errors and non-conservative assumptions in NPSH calculations. In some cases, NPSH reanalyses conducted to support plant modificationr may result in a substantial reduction of margin in available NPSH or a change in the u,ginal design basis cf the plant.

In particular, recent examples indicate that licensees have credited containment overpressure to satisfy NPSH requirements in response to changing plant conditions and errors discovered in earlier NPSH calculations.

RG 1.1 establishes the regulatory position that emergency cure cooling and containment heat removal systems should be designed so that adequate NPSH in provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present before any postulated LOCAs. NRC Standard Review Plan (SRP) 6.2.2, " Containment Heat Removal Systems" (NUREG-0800, Revision 4, dated October 1985) clanfies RG 1.1 by stating that the NPSH analysis should be based on the assumption that the containment pressure equals the vapor pressure of the sump water, in order to ensure that credit is not taken for containment pressurization during the transient.

As part of licensing and Systematic Evaluation Plan reviews, the NRC staff has, in the past, selectively allowed limited credit for a containment pressure that is above the vapor pressure of the sump fluid (i.e., an overpressure) to satisfy NPSH (equirements on a case-by-case basis.

l Reauested Information i

l On the basis of the preceding discussion and examples, addressees are requested to review, l

for each of their respective reactor facilities, the current design-basis analyses used to

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GL 97-04 October 7,1997 Page 7 of 10 determine the available NPSH for the emergency core cooling (including core spray and decay heat removal) imd containment heat removal pumps that meet either of the following criteria:

(1) pumps that take suction from the containment sump or suppression pool following a design-basis LOCA or secondary line break, or (2) pumps used in " piggyback" operation that are necessary for recirculation cooling of 'he reactor core and containment (that is, pu;nps that are supplied uy pumps which take suction directly from the sump or suppressior; pool).

Based on this review, wit' ' 90 days from the date of tMS generic letter, addressees are requested to provide the information outlined below for each of their facilities. New NPSH analyses are neither requested nor required.

1.

Specify the general methodology used to calculate the head loss associated with the ECCS suction strainers.

2.

Identify the required NPSH and the available NPSH.

3.

Specify whether the current design-oasis NPSH analysis differs from the most recent analysis reviewed and approved by the NRC for which a safety evaluation was issued.

4.

Specify whether containment overpressure (i.e., containment pressure above the vapor pressure of the sump or suppression pool fluid) was credited in the calculation of aveilable NPSH. Specify the amount of overpressure needed and the minimum overpressure available.

5.

When containment overpressure is credited in the calculation of available NPSH, confirm that.an appropriate containment pressure analysis was done to stablish the minimum containment pressure.

Reauired Respo ise Within 30 days from the date of this generic letter, each addressee is required to submit a written re:ponse indicating (a) whether or not the requested information will be submitted, and (b) whether or not the requested information will be submitted within the requested time period. Addressees who choose not to submit the requested information, or are unable to submit the information within the requested period, must describe in their response an altemative course of action that is proposed to be taken, including the basis for the acceptability of the proposad attemative.

After reviewing responses to this generic letter, the NRC staff will notify individual addreesees if concems are identified with regard to their facilities.

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Gl. 97-04 October 7,1997 Page 8 of 10 Addressees should submit the required wntien response to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk Washington, D.C. 20555-0001, under oath or affirmation under the provisions of Section 182a, Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f).

Backfit Discussion This generic letter only requests information from addressees under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). The requested information will enable the staff to determine whether addressees' NPSH analyses for the emergency core cooling (including the core spray and decay heat removal) and containment heat removal system pumps conform with the current licensing basis for their respective facilities, inclL 1ng the licensing safety analyses and the principal design enteria which require and/or commit that safety related components and systems be provided to mitigate the consequences of design-basis accidents.

In particular,10 CFR 50.46(a)(1)(i), which addresses the ECCS acceptance criteria for light-water nuclear power reactors, requires in part that the calculated cooling performance of the ECCS following a postulated LOCA conforms to the criteria set forth in 10 CFR 50.46, including provisions for peak cladding temperature and long-term cooling. The potential for loss of adequate NPSH for ECCS pumps, and the cavitation that would result, raises the concem that the ECCS would not be capable of maintaining the peak cladding temperature below acceptable limits, and/or would not be capable of providing core cooling over the duration of postulated accident conditions, as required by 10 CFR 50.46.

Furthermore, the licensing bases of some plants credit the operation nf containment sprays for pressure control as well as for fission product control. The potential for the loss of adequate NPSH for containment spray pumps, and the cavitation that would result, raises the concem that containment spray would not be capable of reducing and maintaining the containment pressure and temperature below design values and would not be capable vf reducing the radiological dose consequences consistent with plants' licensing bases.

Considering the safety significance of removing heat from the containment atmosphere and cooling the reactor core following a design-basis accident, the requested information is needed to venfy addressee compliance with licensing-basis commitments regarding the performance of emergency core cooling (including core spray and decay heat removal) and containment heat removal system pumps. The evaluation required by 10 CFR 50.54(f) to justify this information request is included in the preceding discussion.

Federal Reaister Notification A notice of opportunity for public comment was published in the F:deral Rogister on February 20,1997 (62 FR 7806) to solicit public comments on the draft of this generic letter. A total of 17 comments were received from interested parties, includir.g one industry group, one legal group a' filiated with the nuclear power industry, and two licensees. When I

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GL 97-04 October 7,1997 Page 9 of 10 redundant comments are >:onsidered,12 distinct comments were identified by the staff.

Copies of the sta4 evaluation of these comments have been made available in the NRC Public Document Room.

Paoerwork Reduction Act Statement This gcceric 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 approved by the Office of Management and Budget, approval number 3150-0011, which expires on August 31,2000.

The public reporting burden for this collection of information is estimated to average 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per response, includbg the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needec' and completing and reviewing the collection of information. The NRC 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 proper performance of the functions of the NRC, including whether the information will have practical utility?

2.

Is the estimate of burden accurate?

3.

Is there a way to enhsnce the quality, utility, and clarity of the information to oe collected?

4.

How can the burden of the collection of information be minimized, including the use of automated collection techniques?

Send comments on any aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Mancgement Branch, T-6 F33, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001, and to the Desk Officer, Office of Informatinn and Regulatory Affairs, NEOB-10202 (3150-0011), Office of Management and Budget, Washington, DC 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.

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GL 97-04 October 7,1997 Page 10 of 10 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 managerc Ld h ek W. Roe, Acting Director ision of Reactor Program Management Offica of Nuclear Reactor Regulation Technical contacts: Wiliam O. Long, NRR 301-415-3026 E-mail: wol@nrc. gov Richard M. Lobel, NRR 301 415-2865 E-mail: rml@nrc. gov Lead Project Manager: T.J. Kim, NRR 301-415-1392' E-mail: tjk3@nrc. gov

Attachment:

List of Recently issued NRC Generic Letters i

Attachment GL 97-04 October 7,1997 Page 1 of 1 -

LIST OF RECENTLY ISSUED GENERIC LETTERS Generic

. Date of Letter Subject issuance issued To 97-03 ANNUAL FINANCIAL SURETY 07/09/97 URANIUM RECOVERY LICENSEES UPDATE REQUIREMENTS AND STATE OFFICIALS FOR URANIUM RECOVERY LICENSEES 02 REVISED CONTENTS OF 05/15/97 ALL HOLDERS OF OLs THE MONTHLY OPERATING FOR NPRs, EXCEPT THOSE REPORT WHO HAVE PERMANENTLY CEASED OPERATIONS AND HAVE CERTIFIED THAT FUEL HAS BEEN PER-MANENTLY REMOVED FROM THE REACTOR VESSEL 97-01 DEGRADATION OF CONTROL 04/01/97 ALL HOLDERS OF OLs ROD DRIVE MECHAN 19M FOR PRESSURIZED WATER NOZZLE AND OTHER VESSEL REACTORS,EXCEPT CLOSURE HEAD PENETRATIONS THOSE WHO HAVE PER-MANENTLY CEASED OPERATIONS AND HAVE CERTIFIED THAT FUEL HAS BEEN PERMANENTLY REMOVED FROM THE REACTOR VESSEL 95-06, CHANGES IN THE OPERATOR 02/31/97 ALL HOLDERS OF OLs SUPP.1 LICENSING PROGRAM (EXCEPT THOSE LICENSEES OF PERMANENTLY SHUTDOWN REACTORS WHO ARE NO LONGER REQUIRED TO UTILIZE LICENSED REACTOR OPERATORS) FOR NPRs l

l OL = OPERATING LICENSE CP = CONSTRUCTION PERMIT NPR = NUCLEAR POWER REACTORS l

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POLICY ISSUE (Information)

September 26, 1997 sECY-97-217 FOR:

The Comissioners

.FROM:

L. Joseph Callan Executive Director for Operations

SUBJECT:

PROPOSED GENERIC LETTER ENTITLED " ASSURANCE OF SUFFICIENT NET POSITIVE SUCTION HEAD FOR EMERGENCY CCRE COOLING AND CONTAINMENT HEAT REMOVAL PUMPS." AS REVISED TO REFLECT PUBLIC COMMENTS PURPOSE:

The purpose of this information paper is to inform the Commission, in accordance with the guidance in a memorandum dated December 20. 1991. 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 t

letter.

The subject generic letter requests that addressees submit information relating to the adequacy of the net positive suction head (NPSH) available for emergency core cooling system (ECCS) and containment heat removal pumps when the suction for the pumps is from the containment sump or cuppression pool, or for any pumps necessary for recirculation cooling of the core and containment.

In oarticular, the staff is concerned that some plants have taken credit for containment overpressure to satisfy NPSH requirements, thus possibly creat.ng an inconsistency with the plants' licensing bases. The generic letter, therefore, also requests information regarding each addressee's use of containment overpressure.

The requested information will enable the NRC to determine whether the current NPSH analyses for reactor facilities are consistent with their respective current licensing bases. A copy of the proposed generic letter is attached.

NOTE: To BE MADE PUBLICLY AVAILABLE IN 5 WORKING DAYS FRoM THE DATE oF THis PAPER CONTACT:

William 0. Long. NRR/DSSA (301) 415-3026

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d The Comissioners DISCUSSION:

As a result of recent ins)ection activities, licensee notifications, and licensee event reports, tie NRC staff has identified a safety-significant issue that has generic implications and warrants action by the NRC to ensure that the issue is adequately addressed and resolved.

The issue is that the NPSH available for ECCS (including core spray and decay heat removal) and containment heat removal pumps may not be adequate under all design-basis accident scenarios.

In some cases, this inadequacy may be a result of changes in plant configuration operating procedures, environmental conditions. Or other operating parameters over the life of the plant.

In other cases, a plant's NPSH analysis may not bound all postulated events for a sufficient time, or assumptions used in the analysis may be nonconservative or inconsistent with assumptions and methodologies traditionally considered acceptable by the staff. For example, some licensees have recently discovered that they must take credit for containment overpressure to meet the NPSH requirements of ECCS and containment heat removal pumps.

In the examples the NRC staff is familiar with, the need for crediting this overpressure in NPSH analyses has arisen because of changes in plant configuration and operating conditions, and/or errors in previous NPSH calculations.

The current NPSH analyses (including any corresponding containment pressure analysis) may not be available to the staff in docketed material because some licensees have changed their analyses.

Consequently this generic letter requests that addressees identify whether their current NPSH analysis differs from the most recent analysis that has been approved by the staff.

The generic letter is considered necessary for the following reasons:

(1) there is a high risk significance associated with the potential common-mode failure of ECCS pumps that could result from a sustained loss of NPSH, given an initiating event: (2) all )lants are susceptible to changing plant conditions (e.g.. a change in the p1ysical plant, a change in operating parameters) that could require an NPSH reanalysis and the possiale use of containment overpressure; anu (3) recently, several plants have had especi'lly notable problems with NPSH. On May 9.1997, the licensee for Monticello voluntarily commenced a shutdown after determining that even with credit for containment overpressure, the blockage of the FCCS suction strainers that is postulated to occur during a design-basis loss-of-coolant-accident could degrade the available NPSH to a point at which the ability to adequately cool the reactor core would come into question.

Furthermore, the licensee for the Haddam Neck olant found that 21 psig of containment overpressure is required to meet residual heat removal pump NPSH requirements for the alternate recirculation flowpath, and described the NPSH problem as one of the most safety-significant issues at the plant.

Finally, the licensee for the Maine Yankee plant found that the facility is only able to meet its containment spray pump NPSH requirements for a reduced power level.

The proposed generic letter was originally classified as an " urgent" communication and was transmitted to the Committee To Review Generic Require-ments (CRGR) by a memorandum from Ashok C. Thadani to Edward L. Jordan. dated

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The Commissioners January 6, 1997.

The CRGR was briefed on the proposed generic letter on January 9. 1997 during CRGR meeting number 298.

Following the briefing and incorporation of its comments, the CRGR endorsed the issuance of the 3roposed generic letter on January 17. 1997.

However, on the basis of feedbact from the Deput/ Executive Director for Operations, the generic letter was reclassified as "non-urgent" and was published in the Federal Register on February 20. 1997 (62 FR 7806). to solicit public comments.

Twelve distinct comments were received.

A copy of the si2 *f's resolution of these comments will be made available in the public document room and is attached.

Please note that the staff has subsequently decided to (a) extend the period for submittal of all the requested information to 90 aays in response to comments received since the public comment period ended, and (b) reduce the level of detail of the information being requested.

The ACRS reviewed the proposed generic letter during its 442nd seting on June 12, 1997.

In a letter to Mr. L. J. Callan dated June 17, 1.97. ACRS Chairman Seale stated that the Comittee supports the issuance of the Generic Letter. and commented that (1) the staff needs to define the acceptance criteria for corrective actions. (2) credit for containment overpressure should not be allowed because it may not be available during shutdown and containment bypass sequences, and (3) the '.nspection program needs to be made more effective as the instances of noncompliance identified in the draft generic letter had remained undetected for many years.

Mr. Callan realied to Mr. Seale on August 15. 1997 stating that (1) the staff shares the ACRS' concern about loss of NPSH during bypass sequences. (2) neither the NRC nor industry has considered credit for overpressure during shutdown. (3) for other sequences, the acceptability of credit for overpressure must be addressed on the basis of the probability and consequences of specific bypass scenarios.

(4) the staff will followup on the GL responses with selected inspections based on NPSH margin, and (5) the staff would be pleased to brief the Comittee on the power reactor inspection program.

In a subsequent letter dated September 9. 1997, the ACRS expressed remaining concerns regarding NPSH credit for overpressure, expressed a desire to continue discussion of the subject with the staff, and accepted the offer to be briefed on the inspection program.

The staff intends to issue this generic letter approximately 5 working days after the date of this information paper.

DISTRIBUTION:

Comissione rs

,xj occ oCAA L. < c2ph Callan Executive Director OIG for Operations

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ACRs

Attachment:

1. Proposed Gener'.c Letter 97-XX

2. Staff Resolution of Public Comments cIo

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. The Nuclear Energy Institute (NEI)1 is reviewing a proposed generic letter

-(62 Fed. Reg. 7806 -- February 20,1997) addressing assurance.of sufficient net positive suction head for emergency core cooling'and containment heat removal pumps. Comments on this proposed generic letter will be submitted to your office by March 31,1997.

If you have questions, please contact John Butler at (202) 739 8108.

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- David J. Modeen.

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- H. Dawson, NRC J.~ Kudrick, NRC.

Public Document Room (Project Number 689) 1 NEl is the organization responsible for establishmg umfied nuclear industry policy ou matters affecting the nuclear energy industry, including regulatory aspects of generic operational and technical issues. NEl

= members include all utilitiec licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect / engineering firms, fuel fabrication facilities. materials. licensees,. and other organizations and individuals involved in the nuclear energy industry.

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Subject:

lilinois Power's (IP's) Comments on Proposed Generic Communication; GL 97-XX," Assurance of Sufficient Net Positive Suction llead for Emercency Core Cooline System And Containment Heat Removal Pumns"

Dear Madam or Sir:

This letter is in response to the proposed generic letter that will request addressees to submit analysis and assumptions used to determine net positive suction head (NPSH) for the subject systems. Members of the Clinton Power Station engineering staff have reviewed the proposed GL 97-XX, issued for public comment, and provide the following opinions:

The Generic Letter should clarify that if bounding values are used in the analyses, then time history analyses are not required. It should be sufficient that bounding values are used.

The Generic Letter should clarify that decay heat removal is only required to be analyzed for NPSH concerns when the suction source for the pump is from the suppression pool or reactor building sump.

Item 1(e) proposes to require identifying what quality assurance procedures and engineering program controls were in place when the current NPSH analysis was performed. In our opinion, this requested information is excessive. It should be sufficient to request that licensees ensure their analyses are correct.

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Please consider our comments in completing the final version of the Generic letter.

Sincerely yours, IM.h yr Paul J. Teithorst Director-Licensing FAOiktk rf.

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411 foyetteville Stree. Mall (A1E Raleigh NC 27602 CP&L Letter: PE& RAS-97-034 March 24,1997 Chief. Rules Review and Directives Branch U.S. Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, DC'20555-0001

Subject:

Comments on NRC Proposed Generic Letter on Assurance of Sufficient Net Positive Suetion IIcad for Emergency Core Cooling and Containment IIcat Removal Pumps (62 FR 7806)

Dear Sir or Madam:

Attached please find the comments of Carolina Powei & Light Company (CP&L) to the NRC Proposed Generic Letter (GL) on Assurance of Sufricient Net Positive Suction llead (NPSil) for Emergency Core Cooling (ECCS) and Containment 1leat Removal Pumps. This proposed GL was issued in the February 20,1997 Federal Register (i.e.,62 FR 7806).

Please contact me at (919) 546-6901 if you have questions.

Sincerely, T.D. Walt, Director Operations and Environmental Support IIAS Attachment M

34

- p Page 2 Attachment to CP&L Lctter March 24,1997 PE& RAS-97-034 CP&L Comments on NRC Proposed Generie Letter on Assurance of Sufficient Net Positive Suetion IIcad for Emergency Core Cooling and Containment IIcat Removal Pumps

1. The proposed generic letter requests the Net Positive Suction licad (NPSli) analyses and assumptions for Emergency Core Cooling and Containment Heat Removal pumps. If the analyses are determined not to be in compliance with the Commission's rules and regulations, the affected addressees are expected to take corrective action, as appropriate, in

=

accordance with 10CFR50, Appendix B, to restore the facility to compliance. Raser than providing the NRC with the detai' af the analyses, CP&L suggests that it would be more appropriate for the licensees to con.irm that the NPSH calculations are consistent with the analyses and assumptions in the Final Safety Analysis Report (FSAR). This approach would focus the evaluation onto determining the extent to which the plant configuration agrees with the licensing basis. Therefore, CP&L suggests that the proposed generic letter be revised to have the addressees provide the results of those evaluations and any changes to the Updated Final Safety Knalysis Report, if appropriate, rather than the details of the analyses.

2. If comment No. I above is not incorporated, CP&L suggests that the " Requested Information" section of the proposed generic letter be revised in accordance with the following comments:

Peragraph (1)(d): CP&L suggests that the request for a comparison with the " original licensing-bases analysis" be revised to be a comparist n with the "most current NRC reviewed and approved licensing bases for which a Safety Evaluation was issued." There may have been Safety Evaluations subsequent to the original, and a comparison with potentially out-of-date information serves no purpose and could be misleading and confusing.

Paragraph (3): For completeness, CP&L suggests the addition of the words "and pressure control" after the words heat removal in the first sentence. Both temperature and pressure are important parameters in the calculation of available NPSH, and the subparagraph (3) (c) addresses the pressure issue by requesting information of containment spray use.

Paragraph (3)(a): CP&L suggests that the NRC clarify what is meant by the term

" multipliers" in the sentence: " Identify the heat transfer correlations that were used, and specify whether or not multipliers were used to calculate the transfer of energy to the heat sinks in the containment."

Paragraph (3)(c): CP&L suggests that the NRC consider whether information concerning closed loop cooling systems which exchange heat from the RHR system (or other containment heat removal systems) to the Servi:e Water system should also be requested, for completeness. To omit this could result in an incomplete data base which may then require an additional request for information at some time in the future.

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Page 3-Attachment to CP&L i.etter

- March 24,1997 -

PE& RAS-97-034

CP&L Comments'on NRC Proposed Generic Letter on Assurance of Sufficient Net P6sitive Suetion IIcad for Emergency Core Cooling and Containment lleat Hemoval Pumps cc:

_Mr. J.B. Brady, USNRC Resident inspector - HNP, Unit i Mr. B.B. Desai, USNRC Resident inspector - IIBRSEP, Unit 2 Mr. N.B. Le, USNRC Project Manager - IINP, Unit 1 Ms. B.L. Mozafari, USNRC Project Manager - IIBRSEP, Unit 2 Mr. C.A. Patterson, USNRC Residen' Inspector - BSEP, Units 1 and 2 Mr. D.C. Trimble, Jr., USNRC Project Manager - BSEP, Units I and 2 r

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RE:

Praposed Generic Letter," Assurance of Sufficient Net Pt.sitive Suction llend for Emergency Core Cooling and Containment Heat Removal J2 umps." 62 Fed. Rev. 7.806 (Februarv 20.1997) -

ATfN: Chief, Rules Review and Directives Branch On February 20,1997, the Nuclear Regulatory Commission ("NRC") issued the above-captioned proposed Generic Letter for public comment. Provided below are the comments ofthe Nuclear Utility Backfitting and Reform Group ("NUBARG")# These comments concem the backfitting implications of the proposed Generic Letter.

In the proposed Generic Letter, the NRC reyocsts licensees to review the currem analyses that are used to determine the available net positive suction head ("NPSH") for the emergency core cooling (including core spray and decay heat removal) and containment heat removal pumps which, at arr time following a design-basis accident, take suction from the containment sump or the suppression pool. The proposed Generic Letter cites three examples of nxent findings conceming NPSH issues and states that, while no new NPSH analyses are requested or required, "new NPSH analyses may be warranted if an addressee determines that a facility is not in cor. t e with the Commission's rules and regulations." The requested information is meant to enab..

NRC Staff to determine whether addressees' NPSH analyses for the emergency core cooling and containment heat removal pumps comply and conform with the current licensing basis.

l' l

NUBARG is a consortium of 15 utilities formed in the early 1980s which participated actively in the development of the NRC's backfitting rule (10 C.F.R. {50.109) in 1985, and which has closely monitored the NRC's application of the rule since that time.

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WINSTON & STilAWN U. S. Nuclear Regulatory Commission March 24,1997 Page 2 The proposed Generic Letter lists specific information on the NPSH analyses to be included in the information request; however, it is not clear whether the listed information would have been considered in the original licensing basis of the plant. For example, the list includes an item on the system configurations that were considered in the NPSH analysis for each pump and requests justification for configurations that were not analyzed. If containment overpressure is credited in the analyses, addressees ere asked to give the rationale for the determination of the containment pressure at either a single point or over an extended time period. Whether addressees have modified the analyses or not, the same information is requested and respondents are directed ~

to specify any assumptions.

The "Backfit Discussion" in the proposed Generic Letter discusses a number of General Design Criteria ("GDC") that relate to emergency core cooling or containment heat removal (GDC-35, GDGG8, GDC-16), as well as 10 C.F.R. 650.46," Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," to justify the information requested pursuant to Section 50.54(f). Ilowever, the cited examples of recent NPSH concerns relate to plants with construction permits 2' issued prior to the final rulemaking promulgating the General Design Criterial' and Section 50.46,f as well as prior to the issuance of Safety Guide 1 F which also is mentioned in the proposed Generic Letter. As such, it is not clear how any of the examples relate directly to the backfit discussion, much lessjustify the NRC's information request.

In essence, the proposed Generic Letter requests addressees to verify conformance with certain aspects of the licensing and design bases. Three of the four examples cited by the NRC pertain to concerns with calculational methods that already have been described in detail in NRC Information Notice 96-55, " Inadequate Net Positive Head of Emergency Core Cooling and Containment Heat Removal Pumps Under Design Basis Accident Conditions." In only one of the four examples is the licensing basis discussed; ja, at the Pilgrim plant. In that example, however, the NRC StatY states that "whether or not credit for [ torus] over-pressure is part of Pilgrim's origina!

licensing basis is currently under statT review." In fact, as noted above, the Pilgrim plant obtained a construction permit and was designed prior to the issuance of Safety Guide 1. The proposed Generic Letter does not discuss guidance prior to November 1970 for the NPSH assumptions. Thus, the cited example is of questionable relevance.

F The plants discussed in the proposed Generic Letter and the date of their construction permits are as follows: Haddam Neck, May 26,1964; Maine Yankee, October 21,1968; Pilgrim, August 26,1968; and Crystal River, Unit 3, September 25,1968.

F 10 C.F.R. Part 50, Appendix A. 36 Fed. Reg. 3,255 (February 21,1971).

f 39 Fed. Reg.1,001 (January 4,1974).

F Safety Guide 1 (Regulatory Guide 1.1)," Net Positive Suction Head fer Emergency Core Cooling and Containment Heat Removal Sys'em Pumps," November 2,1970.

+

WINSTON & STRAWN e

U. S. Nuclear Regulatory Commission -

March 24,1997 Page 3 In response to actions associated with NRC Bulletin 96-03," Potential Plugging of

- Emergency Core Cooling Suction Strainers by Debris in Doiling-Water Reactors," most boiling-water reactor licensees will find it necessary to review their NPSH calculations for new strainer installations. Mditionally, as much as the issue relates to conformance with current licensing and design bases, the NRC informed licensees in its revised enforcement guidance that discrepancies between the licensing basis and the actual plant configuration will be considered for enforcement action.F The "Backfit Discussion"in the proposed Generic Letter does not discuss provisions of 10 C.F.R. {50.109, but rather provides "[t]he evaluation required by 10 C.F.R. {50.54(0 to justify this information mjuest." We believe that this type of request is unwarranted and that the Staff has not shown that the burden on licensees is justified.

In the Statement of Considerations accompanying the 1985 amendments to Section 50.54(f), the NRC states that:

If extensive effort is reasonably anticipated, the request will be evaluated to detennine whether the burden imposed by the information request isjustiiled in view of the potential safety significance of the issue to be addressed.... Requests for information to determine compliance with existing facility requirements... usually are not made pursuant to {50.54(f).... The amendment of {50.54(f) should be read as indicating a strong concem on the part of the Commission that extensive information requests be carefully scrutinized by staff management prior to initiating such requests. The Commission recognizes that there may be instances where it is not clear whether a backfit will follow an infonnation request. Those cases shauld be n. solved in favor of analysis:Z' We believe this and the language of the rule itselfindicate the Commission's original intent that Section 50.54(f) be used only for the most significant issues when the Commission n. ist determine whether or not the license of a facility "should be modified, suspended, or revoked."F We recommend that the NRC not issue the proposed Generic Letter until completion ofa backfitting analysis pun,aant to 10 CFR {50.109. Absent the requisite backfitting analysis, the Staff cannot justify the need for the information and any new requirements imposed on licensees through a new interpretation of ple.nt licensing and design bases (s&, imposition of General Design Criteria to pre-GDC-licensed plants). The backfitting analysis should include justification for applying the request to all plants. Altematively, if the Staff believes that it has additional F

Sss 61 Fed. Reg. 54,461 (October 18,1996).

I' 50 Fed. Reg. 38,112 (September 20,1985).

F-10 C.F.R. {50.54(f).

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WINSTON & STRAWN2 i U. S. Nuclear Regulatory Commission -

March 24,-1997 L Page 4 information or msights useful to licensees, a supplement to Information Notice 96-55 could be -

~ issued rather than the proposed Generic Letter, or the concerns could be addressed through.

rulemaking.

Sincerely,'

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-A Ka S M. Sutton Daniel F. Stenger.

Counsel for Nuclear Utility Ba:kfitting and Reform Group -

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Mr. David L. Meyer, Chief g y~ -

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Washington,~ DC 20555-0001 E

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SUBJECT:

Proposed Generic Communication: Assurance of Sufficient Nef Positive Suction Head for Emergency Core Cooling and Containment Hent Removal Pumps (62 Fed. Reg. 7806 - February 20,1997)

Dear Mr. Meyer:

Enclosed are comments submitted on behalf of the nuclear power industry by the Nuclear Energy Institute (NEI).1 These comments are in response to the February 20,1997, Federal Register notice concerning the subject proposed generic letter.

We appreciate the opportunity to comment on this proposed generic letter.

Please direct any questions on our comments to John Butler at (202) 739-8108.

Sincerely, OWh v'

J, Modeen JCB/cip Enclosure -

c:

. Mr. Stewart L. Magruder, NRC Mr. Howard (Jack) Dawson, NRC Mr. John A. Kudrick, NRC Public Document Room (Project No. 689)

1. NEI is the organization responsible for establishing unified nuclear industry pol;cf on matters affecting the nuclear energy industry including regulatory aspects of generic operational and technical issues. NEI members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant heigners major architect / engineering Srms, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry.

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Industry Comments on Proposed Generic Communication:

Assurance of Sufficient Net Positive Suction Head (NPSH) for Emergency Core Cooling and Containment Heat Removal Pumps -

(62 Fed. Reg. 7806 - February 20,1997) 1.

Comment:

Tho' proposed generic letter requests that addreues provide information on NPSH analyses within 60 days from the dato of the generic letter. Recent generic letters requesting a similar level ofinformation have provided a response period ranging from 90 days to 180 days. The abbreviated response period of the proposed generic letter, if maintained, will necessitate a re-prioritization oflicensee planned activities. The issues identified in the proposed geperic letter do not warrant such a short response period. The proposed generic letter requests the collection and submittal of a considerable amount ofinformation. Sufficient time should be allowed to prepare the information requested by the proposed generic letter.

The safety issues in the proposed generic letter were previously identified through NRC Information Notice (IN) 96 55," Inadequate Net Positive Suction Head of Emergency Core Cooling and Containment Heat Removal Pumps Under Design Basis Accident Conditions." Licensees were asked to review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. Licensee evaluations of IN 96-55 would have already prompted any required short-term actions.

Recuested Action:

The response period should be increased to at least 90 days to allow sufficient i

time for the collection and preparation of the requested information.

2.

Comment:

Item (1) under the Requested Information section of the proposed generic letter states:

" Provide the NPSH analysis and assumptions for each pump, and. in l

particular...."

This request continues with a specification of the "particular" information that is being sought by the NRC staff.

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--Thei resence of the second "and"in the alwve request makes it unclear whether.

submittal of the "particular" information will fully satisfy the request or -

- whether additional infctmation on "NPSH analysis and assumptions" is L requested. Forwarding of compmaensive analysis packages would appear to be more than what is intended, at least from reading the Discussion section.

Reauested Action:-

I Please provide clarification on the requested information by identifying whether

- a response to the specific requests (i.e., " particulars") will satisfy the

. information request. If not, please identify what parts of the NPSH analysis are needed.

3.

Comrnent:

Item (1) ungler the Requested Information section of the proposed generic letter requests " analysis and assumptions for each numo...."

- Plant analyses are often performed with the recognition that groups of.

redundant pumps (e.g., High Preasure Safety Injection, Containment Spray) may be in operation. Information on individual pump operation may not be available. Plants may also have pump configurations in which one or more pumps do not take direct suction from the containment sump or the suppression-pool, but are supplied by a pump.

Heauested Action:

The'information request should be modified to allow licensees to provide information for groups of pumps when applicable to the analysis. Also, please clarify whether the information request is limited to pumps that take suction directly from the containment sump or suppression pool.

.4.

Comment:

. Item (1)(a) under the Requested Information section of the proposed generic letter states:

s "Specify, as a function of time. the required NPSH and the available NPSH."

.The analyses used to determine NPSH make use of conservative assumptions to -

E define required and available NPSH valu3s. The use of these assumptioas can result in a single, maximum required NPSH and a single, minimum avriilable

NPSH..These are compared to ensure that adequate NPSH is availab'.e

^

throughout the required time frame. To provide NPSH values as a fr.netion of ling, might require the performance of a new separate analysis.

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Enuested Action:

The information request should be modified to acknowledge the submittal of bounding NPSH values as an acceptable response to this request.

5.

flornment:

Item (1)(d) under the Requested Information section of the proposed generic letter states:

"Specify if the current licensing basis NPSH analysis is different from the oricinal licensing basis analysis,..."

The originallicensing basis analysis might have been replaced by a subsequent analysis which has been reviewed and approved by the NRC. Where this is the case, the alpve request could potentially result in a comparison to out-of date information.

Reauested Action:

The generic letter should clarify the above request to specify whether the original licensing-basis analysis or the most current NRC reviewed and approved analysis should be used.

G.

Comment:

Item (1)(c) under the Requested Information section of the proposed generic letter states:

"Specify any quality assurance procedures and engineering program controls in place when the current NPSH analysis was performed."

This request in its current form is very broad and appears to be inconsistent with the stated purpose of the information request, which is "... to determine if the NPSH analyses for reactor facilities are consistent with their respective current licensing basis."

hauested Actioru NRC staff should review the purpose of this request and either remove the request or provide a clearer specification of the information requested and the basis for the request.

3

e 7;

Conunent:

In at least one instance, a licensee has submitted written NPSH analysis -

a documentation to the NRC and has received an NRC safety evaluation report for the same.

Reauested Action:

The generic letter should be modified to allow reference to previously submitted and accepted NPSH analysis information.

8.

Comment:

The focus of the proposed generic letter, as identified in the Requested Information section, is on NPSH analyses for events in which the emergency core coolingeor containment heat removal pumps take suction from the containment sump or the suppression' pool. This focus excludes analyses for secondary system pipe breaks as they do not result in pump suction from the containment sump or the suppression pool.

The Background section of the proposed generic letter states:

"This generic letter applies only to ECCS and containment heat removal pumps that take suction from the containment sump or suppression pool following a loss of-coolant accident (LOCA) or secondary line break."

The Requested Information section of the proposed generic letter, under item (1)(b) states;

" Identify the postulated pipe breaks that were analyzed if a spectrum of primary and secondary system pipe break sizes and locations was considered...."

Reauested action:

Please provide clarification that the analysis information requested in the proposed generic letter is' limited to those time frames during which the emergency core cooling and containment heat removal pumps are taking suction from the containment sump or the suppression pool. It is also recommended that the words "or secondary line break" be deleted from the Background section and that "and secondary" be deleted from the Requested Information section.

4

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STAFF RESOLUTION OF PUBLIC COMMENTS RECEIVED ON DRAFT GENERIC LETTER ENTITLED

" ASSURANCE OF S'FFICIENT NET POSITIVE SUCT:0N HEAD FOR EMERGENCY CORE COOLING J

AND CONTAINMENT HEAT REMOVAL PUMPS" (62 FR 786. February 20. 1997)

The staff received a total of 17 comments on the draft generic letter entitled

" Assurance of Sufficient Net Positive Suction Head For Emergency Core Cooling and Containment Heat Removal Pumps." Of these. 5 were redundant, leaving 12 distinct comments licusing primarily on clarification of the information requested in the generic letter.

The majority of the comments came directly from utilities or from the Nuclear Energy Institute (NEI), and one comment came from the Nuclear Utility Backfitting and Reform Group (NUBARG) associated with the law offices of Winston and Strawn.

The following discussion provides the coments received and the NRC staff's response to these comments.

The staff's responses st'te clearly whether, and low, the generic let'er was l

revised to reflect a particular comment.

I.

Carolina Power and Licht Comoany (CP&L)

Comment 1:

"The proposed generic letter requests the Net Positive Suction Head (NPSH) analyses and assumptions for Emergency Core Cooling and Containment Heat Removal pumps. If the analyses are determined not to be in compliance with the Commission's rules and regulations, the affected addressees are expected to take corrective action. as aopropriate. in accordance with 10 CFR 50. Appendix B. to restore the facility to compliance.

Rather than providing the NRC with the details of the analyses. CP&L suggests that it would be more appropriate for the licensees to confirm that the NPSH calculations are consistent with the analyses and assumptions in the Final Safety Analysis Report (FSAR). This approach would focus the evaluation onto determining the extent to which the plant configuration agrees with the licensing basis.

Therefore. CP&L suggests th't the proposed generic letter be revised to have the addressees provide the results of those evaluations and any changes to the Updated Final Safety Analysis Report. if appropriate.

rather than the details of the analyses."

Response: The proposed gener~; letter has been revised in a manner which significantly reflects this comment.

Comment 2: "If coment No.1 above is not incorporated. CP&L suggests that the

' Requested Information' section of the proposed generic letter be revised in accordance with the following comments:"

a)

" Paragraph (1) (d):

CP&L suggests that the request for a comparison with the ' original licensing-bases analysis' be revised to be a comparison with the

'most current NRC reviewed and approved licensing bases for which a Safety Evaluation was issued.'

There may have been Safety Evaluations subsequent to the original. and a comparison with potentially out-of-date information serves no purpose and could be misleading and confusing."

Response

The staff agrees with this comment, and has revised the generic Attachment 2

~

letter to request that addressees compare the design basis NPSH analysis with the most current NRC-reviewed and approved analysis fo,1 which a safety evaluation was issued.

b)

" Paragraph (3):

For completeness. CP&L suggests the addition of the words

'and pressure control * -after the words ' heat removal' in the first sentence.

Both~ temperature and 3ressure are impcriant parameters in the calculation of available NPSH. and tie subparagraph (3) (c) addresses the pressure issue by requesting information of containment spray use."

ResDonse: The propesee generic letter has been significantly revised.

Tnat language.has been deleted from the " Requested Information" section.

c)

" Paragraph (3) (a):

CP&L suggests that the NRC clarify what is meant by the term ' multipliers

• in the sentence:

' Identify the heat transfer correlations that were used, and specify whether or not multipliers were used to calculate the transfer of energy to the heat sinks in the containment.'"

Resoonse:

The proposed generic letter has been significantly revised.

The

" multipliers" request has been deleted from the " Requested Information"

-section.

d) " Paragraph (3) (c):

CP&L suggests that the NRC consider whether information concerning closed loop cooling systems-which exchange heat from the RHR system (or other containment heat removal systems) to the Service Water system should also be requested. for completeness.

To omit this would result in an incomplete data base which may then require an additional request for information at some time in tne future."

Resoonse: The proposed generic letter has been significantly revised.

That language has been deleted from the " Requested Information" section.

II.

Illinois Power Comment 1:

"The Generic Letter should clarify that if bounding values are used-in the analyses, then time history analyses are not required.

It should be sufficient that bounding values are used."

Resoonse: The proposed generic letter has been revised to reduce the level of detail of information being requested.

Addressees will not be requested to furnish information with this level of detail in their initial responses.

However, after reviewing the initial information, the staff may find it necessary to request additional information, at which time this comment will be considered.

Comment 2:

"The Generic Letter should clarify that decay heat removal is only required to be analyzed for NPSH concerns when the suction source for the pump is from the suppression pool or reactor building sump."

Resoonse:

The focus of the staff's concern relates to the recirculation phase and to the credit for containment overpressure. Although NPSH for pumps taking suction from a cold storage tank (e g.

refueling water storage tank or Attachment 2

O i

e cond?nsate storage tank) must be assured, that concern is not encompassed by the proposed generic letter.

The staff has revised the generic letter to clarify that the scope of information requested applies only to emergency core cooling system (ECCS) and containment heat removal pumps that meet either of the following criteria:

1) pumps that take suction from the cantainment sump or suppression pool following a design-basis LOCA or secondary line break, or

2) pumps used in " piggyback" operation that are necessary for recirculation cooling of the reactor core and containment (that is, pumps that are supplied by pumps which take suction directly from the sump or suppression pool) (See Comment 3 from NEI).

Commant 3:

" Item 1[1](e) proposes to require identifying what quality assurance procedures and engineering program controls were in place when the current NPSH analysis was performed. In our opinion, this reques+ed information is excessive. It should be sufficient to request that licenst.es ensure their analyses are correct."

Resoonse: The staff agrees and has revised the proposed generic letter accordingly.

III.

Nuclear Eneray Institute (NEI)

Comment 1:

"The proposed generic letter requests that addressees provide information on NPSH analyses within 60 days from the date of the generic letter. Recent generic letters requesting a similar level of information have provided a response period ranging from 90 days to 180 days. The abbreviated response period of the proposed generic letter, if maintained, will necessitate a re-prioritization of licensee-planned activities. The issues identified in the proposed generic letter do not warrant such a short response period. The proposed generic letter requests the collection and submittal of a considerable amount of information. Sufficient time should be allowed to prepare the information requested by the proposed generic letter."

"The safety issues in the proposed generic letter were previously identified through NRC Information Notice (IN) 96-55

' Inadequate Net Positive Suction Head of Emergency Core Cooling and Containment Heat Removal Pumps Under Design Basis Accident Conditions.

Licensees were asked to review the information for applicability to their facilities and consider actions. as appropriate, to avoid similar prob' ems. Licensee evaluations of IN 96-55 would have already prompted any required short-term actions."

"The response period should be increased to at least 90 days to allow sufficient time for the collection and preparation of the requested information."

Resoonse: The proposed generic 'etter has been revised to specify a 90-day response period and the scope of information requested has been reduced.

Comment 2:

" Item (1) under the Requested Information section of the proposed Attachment 2 l

)

_ generic letter states:

' Provide the NPSH analysis and assumptions for each pump, mgL in particular This request continues with a specification of the "particular" information that is being taught by the NRC staff."

."The presence of the second 'and' in the above request makes it unclear whether submittal of the "particular" information will fully satisfy the request or whether additional information on 'NPSH analysis and assumptions' is requested. Forwarding of comprehensive analysis >ackages would appear to be more than what is intended..at least from reading tie Discussion section.

Please provide clarification on the requested information by identifying whether a response to the specific requests (i.e.

" particulars") will satisfy the information request. If not, please identify what parts of the NPSH analysis are needed."

Resoonse:

The proposed generic letter has been revised to reduce the level of detall of information being requested.

Addresses will not be requested to furnish information with this level of detail in their initial response.

However, after reviewing the initial information, the staff may find it necessary to request additional information.

~ Comment 3:

" Item (1) under the Requested Information section of the proposed generic. letter requests ' analysis and assumptions for each oumo...'"

" Plant analyses are often performed with the recognition that groups of redundant pumps (e.g., Higi Pressure Safety Injection. Containment Spray) may be in operation. Information on individual pump operation may r.ot be available. Plants may also have pump configurations in which one or more pumps do not take direct suction from the containment sump or the suppression pool, but are supplied by a pump."

"The information request should be modified to allow licensees to provide information for groups of pumps when applicable to the analysis. Also, please clarif whether the information request is linited to p mps that take suctio.,

direct y from the containment sump or suppression pool Resoonse: The proposed generic letter has been revised to reduce the level of detail of information being requested.

Addresses will not be requested to furnish information of this level of detail in their initial response.

However, after reviewing the initial-information, the staff may find it necessary to request additional infon.1ation.

Comment 4:

" Item-(1)(a) under the Requested Information section of the propcsed generic letter states:

'Specify as a function of time. the required NPSH and the av

. ale NPSH.

• The analyses used to determine NPSH make use of conservative assumptions to Attachment 2

s define required and available NPSH values. The use of these assumptions can

. result in a single, maximum required NPSH and a single, minimum available NPSH.

These are compared to ensure that adequate NPSH is 6vailable throughout the required-time frame. To provide _NPSH values as a function of time might require the performance of a new separate analysis."

"The information request should be modifleo_to acknowledge the submittal of bounding NPSH values as an acceptable response to this request."

Resoonse: See the staff's response to Comment 1 from Illinois Power.

Comment 5:

" Item (1)(d) under the Requested Information section of the proposed generic letter states:

'Specify if the current licensing basis NPSH analysis is different from the cricin61 licensing-basis analysis..

The original licensing-basis analysis might have been replaced by a subsequent analysis which has been reviewed and approved by the NRC. Where this is the case, the above request could potentially result in a comparison to out-of date information."

"The generic letter should clarify the above request to specify whether the original licensing-basis analysis or the most current NRC reviewed and approved analysis should be used."

Resoonse: See the staff's response to Comment 2a from Carolina Power and Light Company.

Comment 6:

" Item (1)(e) under the Requested Information section of the proposed generic letter states:

'Specify any quality assurance procedures and engineering program controls in place when the current NPSH analysis was performed.'

This request in its current form is very broad and Jppears to be inconsistent with the stated purpose of the information request, which is

• to determine if the NPSH analyse 2 for reactor facilities are consistent with their respective current licensing basis.'"

"NRC staff should review the purpose of this request and either remove the request or provide a clearer specification of the information requested and the basis for the request."

Resoonse:

See the staff's response to Comment 3 from Illinois Power.

Comment 7:

"In at-least one instance, a licensee has submitted written NPSH analysis documentation to the NRC and has received an NRC safety evaluation report for the same."

"The generic letter should be modified to allow reference to previously submitted and accepted NPSH analysis information." Attachment 2

4 Ersponsg:

If addressees know that the staff is already in possession of any requested information, they may state that fact in their response to the generic letter.

The staff has revised the generic letter to reflect the response to this comment.

4Lmwtn,Lf:

"The focus of the proposed aeneric letter, as identified in the P

Requested Information section is on NPdH analyses for eventL in which the energency core cooling or containment heat removal pumps take suction from the containment sump or the su)pression pool. This focus excludcs analyses for secondary system pipe breats as they do not result in pump suction from the containment sump or the suppression pool."

The Background section of the proposed generic letter states:

'This generic letter applies only to ECLS and containn.ent heat removal pumps

-that take suction from the containment sump or suparession poc following a loss-of coolant accident (LUCA) or secondary line 3reak '"

~"Ee Requested Information section of the proposed generic letter, under item (1)(b) states:

' Identify the postulated pipc breaks that were analyzed if a spectrum of primary and secondarv system pipe break sizes and locations was considered "Please provide clarification that the analysis information requested in the proposed generic letter is limited to those time frames during which the emergency core cooling and containment heat removal pumps are taking suction from the containment sump or the suppression pool. It is also recommended that the words 'or secondary line break' be deleted from the Backgrour.d section and that *and secondary' be deleted from the Requested Information section "

Response

The information in the comment regarding the p.:ap suction source has always been the intent of the generic letter.

Consequently, the staff has revised the generic letter to clarify that the information request applies only to ECCS and containment heat removal pumps that take suction from the containment sump or suppression cool following a LOCA or are required for l

recirculation cooling of the reactor core and containment (See the staff's response to Coment 2 from Illinois Power).

The concent also addressed the issue of secondary line breaks.

If, in the event of a steam or feedwater line break, containment spray pumps woulo nave to eventually operate in a recirculation mode, adequacy of NPSH must be analyzed.

The staff thus has no reason to exclude secondary breaks from the scope of information requested. Consequently, the staff has not revised the generic letter with regard to this requirement.

ly Winston and Strawn - Nuclear Utility Backfittina and Reform Grouc 1hUSARG)

Comment 1:

"We recommend that the NRC not issue the proposed Generic Letter until completion of a backfitting analysis pursuant to 10 CFR 550.109. Absent Attachment 2

1..

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the requisite backfitting analysis, the Staff cannot justify the need for the information and any new requirements imposed on licensees through a new interpretation of plant licensing and design bases (e.g., im>osition of General Design Criteria to pre GDC licenseo plants). The baccfitting analysis should include justification for applying the request to all plants.

Alternatively. if the Staff believes that it has additional information or insights useful to licensees, a supplement to Information Notice 96 55 could be issued rather thn the proposed Generic Letter, or the concerns could be I

addressed through rulemaking "

Resoonse:

It is not the intent of the generic letter to impose new require-ments or new interpretations of plant licensing and design bases on licensees.

Rather. the intent of the generic letter is to recuest information appropriate to the staff's recognition of licensees' increasec reliance on containment overpressure, as a result of errors in the NPSH calculation or changes in plant design, to sat'sfy NPSH requirements.

The generic letter constitutes a request for 1rformation only.

The comment specifically addressed application of General Design Criteria (GDC) to plants licensM before the promulgation of those criteria (" pre-GDC" plants). The staff riotes that " pre-GDC" plants were reviewed and approved using criteria that were essentially the same as the GDC.

In the Statement of Considerations for the proposed GDC the Commission stated that "these General Design Criteria would not add any new requirements, but are intended to describe more clearly present Commission requirements to assist applicants in preparing applications." This view was reiterated in SECY-92-223, " Resolution of Jeviations identified During the Systematic Evaluation Program." dated June 19, 1992.

Finally, the staf f notes that the introduction to 10 CFR Part 50. Appendix A states that the GDC " establish minimum requirements for the principal design criteria for water-cooled nuclear power )lants similar in design and location to plants for which construction permits lave been issued by the Commission" (emplasis added).

This reinforces the view that the GDC were not completely new requirements, but rather represented a codification of existing NRC review and approval practices.

Therefore, mention of the GDC in the backfit discussion of the generic letter does not impose new requirements on any licensed plant.

Rather, the GDC simply formalize previously existing licensi m requirements and practices.

With regard to the mention of 10 CFR 50.46 in the bac': fit oiscussion in the aeneric letter, the staff notes that addressees are equired to meet either 0 CFR 50 J6. GDC 35, or both. The only plants that may not need to meet 10

( H 50.46 would be those plants without Zircaloy fuel cladding.

In such cases, the particular plant would need to meet criteria very similar to those in 10 CFR 50 A6. and would still need to comply with GDC 35, which specifies that a facility must have a.rystem to' provide

" abundant emergency core cooling."

Finally.10 CFR 50.54(f) states that. "Except for information sought to verify licensee compliance with the current licensing basis for that facility, the NRC must prepare the reason or reasons for each information request prior to

.. issuance to ensure that the burden to be imposed on respondents is justified in view of the potential safety significance of the issue to be addressed in 7-4

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l the requested-information." The request for information is being made to ensure that licensees are in compliance with their current licensing bases.

consistent with 50.54(f).

Because the generic letter makes no attempt to impose new requirements or new l

interpretations of plant licensing and design bases as-discussed above, and i

because the information is being requested.n accordance with 10 CFR 50.54(f).

i a backfitting analysis pursuant to-10 CFR 50.109 is not riecessary.

Therefore.

the staf f has not revised the generic letter to reflect this comment, j

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- Attachment 2-

F CRGR REVIEW PACKAGE STAFF RESPONSES TO QUESTIONS IN SECTION IV.B 0F lHE CHARTER DOCUMENT FOR THE COMMITTEE TO REVIEW GENERIC REQUIREMENTS This document constitutes an attachment submitted to the Committee To Review Generic Requirements (CRGR) by the staff of the Office of Nuclear Reactor Regulation (NRR) in association with the memorandum requesting review and endorsement of the proposed generic letter entitled " Assurance of Suff1cient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps." as revised to reflect public comments.

As such, this attachment 3 resents the staff's responses to the questions contained in Section IV.s of the CRGR Charter, dated April 19. 1996.

These responses document the staf f's justification for the responses regulated by Title 10.

Sectior 50.54(f). of the Code of federal Regulations (10 CFR 50.54(f)).

PROPOSED ACTION:

Issue a generic letter requiring addressees to submit the results and certain key issum)tions of the current design-basis analysis used to show tlat suffi::1ent net positive suction head (NPSH) would exist for emergency core cooling and containment heat removal pumps under design-basis acci-dent conditions.

In particular, licensees are requested to submit details of the containment pressure response analysis used to determine containment overpressure if containment overpressure is credited ior NPSH.

The draft generic letter was issued for. and received, public comment, after being published in the Federal Register (62 FR 786).

CATEGORY:

42 RESPONSE TO REOUIREMENTS FOR CONTENT OF PACKAGE SUBMITTED FOR CRGR REVIEW (i)

Tile proposed generic requirement or staff position as it is proposed to be sent out to licensees. Where the objective or intended result of a proposed generic requirement or staff position can be achieved 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 licensee how the objective or result is to be attained.

W+%4n C0 days--of-the date of thi" generic letter addressee; are mered to catWt to the MC..ritten-anformatice stating the results anc key assumpHens of the currect 'icensing basis analysis used to determine the net p0;itive cuction head available for each emergency ese-eesh -and contem4et--remo,'al system pump.Mich takes 3

suc44sn frc'" the recirculaHen su"tp or suppression pool

I.

2 Addressees are requested to review. for each of their respective reactor facilities, the current design basis analyses used to determine the available NPSH for the emergency core cooling (including core spray and decay heat' removal) and containment heat removal pumps that meet either of the following criteria:

1)-pumps that-take suction from the containment sump or suppression pool following a design basis LOCA or secondary line break, or-

2) pumps used in " piggyback" operation that al necessary for recirculation cooling of. the reactor core and containment (that is.

pumps that are supplied by pumps.which take suction directly.from the sump;or, suppression pool).

Bated on this review. within 60 days from~the date of this generic letter, addressees are requested to provide the information outlined below for each of their facilities. New NPSH analyses are neither requested nor required.

To the extent practical, the use of a tabular format is acceptable in ) resenting the information.

if an addressee references previously su)mitted material to satisfy any part of the information request, the addressee should also submit an accompanying outline, or "roadmap " giving references for and the context of the previously submitted material; The following requested information may be submitted for each of the aforementioned pumps, or information for groups of pumps (e.g., similar pumps that operate in parallel) is acceptable if this is the expected mode of pump operation in the desigr basis analysis for the particular pumps.

(1)

Provide the design-basis NpSH anO ysis used to determine that adequate NPSH is provided for-1ong-term retirculation capability following a t0CA. At a minimum, the following information should be included: and assumptions for cach pump--and. ir, particular--

(a)

S)ecify the working equation used to calculate the availabic NPSH, tie equations used to calculate piping line, exit, and entrance losses, and the numerical values of these losses.

Specify the general methodology used to calculate the head loss associated with the ECCS suction strainers and state the general clogging one strainer completely blocked). = Specify assumptions made (e.g,82 was used to determine the blockage of the if Regulatory Guide 1.

suction straincrs, and specify which revision was used.

(ab) Specify. as a function of t%e-the required NPSH and the available NPSH.

If the current design basis analysis is bounding (rather than time dependent), provide justification as to why the analysis is considered bounding throughout the post-LOCA recirculation period.

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. (bc) identify Consistent with the requirements of 10 CFR 50;46i specify

-the various postulated )ipe breaks and locations that were analyzed to determine tie das19n. basis for NPSH. if a pectru of priaryandsesendar!"Hanalysis.syste p pe break size; and location xa co'cidered in-the NP y and decay

.(ed).Specifytheemergencycorecooling(includingcoresprjigurations heat removal) and containment heat removal system ccm alignments (and associated flow rates) that were considered in the NPSH analysis 'for each pump or group of sumps. Also identify and justify any_ credible system alignments t1at which co-figuraHess were not analyzed.

(de) Specify whether the current licen 4n" design basis NPSH analysis differs from themostrecentianalysisreviewedandaprovedby the NRC and:for which a safety evaluation was issued? phe eriginal licensing basis analysis. and (ef) S)ecify Whether the any quality assurance procedures applied to tie NPSH analysis ~are of the same level as those typically applied to other design basis analyses, or -if, the NPSH analysis' received =a level of quality assurance greater or less than that typical for otherdesign-basisanalyPSHanalysi;na;perfored.ses, p d engi cering pr place when the current (2)

For each pump or group of pumps, specify whether containment overpressure. (i.e.. containment pressure above the vapor pressure of the sump or suppression pool fluid) was credited in the calculation of available NPSH.

Specify the amount of over)ressure needed and the minimum overpressure available.

Indicate w1 ether the overpressure was determined from the containment pressure at a single point in time, or if the calculation considered the containment pressure profile over an extended period.

In the latter case. state the duration of the extended period and give the ritionale for choosing this time period; if the calculation considered containment pressure at only a single point in time. state time and give the ratior. ale for selecting this_ point.

(3)

When containment overpressure is credited in the calculation of available NPSH specify the containment-atmosphera heat removal and pressure control assumptions that were used in ti.'t containment response analysis to determine the minimum containment overpressure available, in particular, provide the following related information:

(a)

Specifylthe' general methodology used to' calculate:the.'c~ontainment pressure (eig,,atmechanistic calculation-conducted with a computer code, an>equiliorium calculation which: assumes the containment pressure follows the suppression pooF temperature (if.

applicable)e etc.). '

4 (ab) Specify the correlations that were used to calculate the heat transfer from the containmant atmosphere to the heat sinks in containment. Also specif whether any multi)licative constants were applied to the corre{.ations to modify t1e calculated heat transfer-(e.g.. a multiplier of 1.2 on the Uchida heat transfer correlation). whether or not multipliers were 'used to calculate the transfer of energy to the heat stnt in the containment-(bc) Specify how many trains of containment spray were assumed to be operating and whether a minimum, maximum, or intermediate value of spray flow was assumed.

(ed) Specify the method and rationale used in the NPSH analysis to select the service water temperatures for the heat exchangers that remove energy from the containment atmosphere and the suppression pool, in boiling water reactors (BWRs) or the containment. sump, in pressurized water reactors (PWRs). Also specify any special assumptions made concerning bmt transfer across the heat exchangers (e.g.. effect of fouling on f. eat transfer. effect of plugged tubes, etc.).

For exam)le.'in PWRs, considerlany heat exchangers that interact with t1e containment sump via the containment spray and RHR pumps, and consider the heat-transfer assumptions for the containment fan coolers, if applicable; For BWRs, consider any heat exchangers that interact with the suppression pool, either in suppression pool cooling'or containment spray mode.

(de) Specify the total number of containment fan coolers at the plant, and specify how many fan coolers were assume:' to be operating.

(ii)

Draft staff papers or other underlying staff documents supporting the 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 reference material for his or her use.)

On October 22. 1996 ' a NRC staff issued Information Notice 96-55.

" Inadequate Net P ive Suction Head for Emergency Core Cooling and Containment Heat

. oval Pumps Under Design Basis Accident Conditions."

in response to events at the Haddam Neck and Crystal River nuclear plants and reports from the licensee for Maine Yankee regarding potentially inadequate NPSH for the emergency core cooling and containment heat removal pumps in the recirculation mode of operation.

5-NRC Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling System and Containment Heat Removal System Pumps." dated November 2, 1970 states the NRC's regulatory position regarding certain assumptions that rnay be used in the analysis to determine the NPSH available for the subject pumps.

In particular. licensees should assume the maximum temperature of the pumped fluid and no increase in containment pressure from that present prior to a postulated LOCA.

(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, in accordance with NRC regulations and licensing commitments, safety-related components are expected to be capable of performing their required safety functions under design-basis accident conditions.

Ali addressees are required to review the current design-basis analysis used to determine the NPSH availabP for their emergency core cooling and containment heat removal system pum)s.

In addressees. addressees are requested to provide to the staff t1e results and certain key atsumptions used in the analysis.

This generic letter neither increases nor reduces existing requirements or staff oositions.

Rather. it requests information to ensure that existing 1RC requirements are met.

(iv) The proposed method of implementation with the concurrence (and any coments) of OGC on the method proposed.

The concurrence of affected program offices or an explanation of any nonconcurrences.

The proposed method of implementation is the issuance of a generic letter.

OGC has no legal objection to this proposed generic letter.

(v)

Regulatory analyses conforming to the directives and guidance 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 documented evaluation is required as discussed in IV.B,(ix).)

This item is not applicable to this proposed staff action.

(vi)

Identification of the category of reactor plants to which the generic requirement or staff position is to apply (that 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 constructior,, 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.).

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6-This proposed generic letter applies to all nuclear power plants, except those who have certified to a permanent cessation of operations.

(vii) For backfits other than compliance or adequate protection backfits, a backfit analysis as defined in 10 CFR 50.109.

The backfit analysis i

shall include, for each category of reactor plants, an evaluation which l

demonstrates how the action should be prioritized and scheduled in light I

The backfit analysis shall.

of other ongoing regulatory activities.

document for consideration information available concerning any of the 1

following factors as may be appropriate and any other information relevant and material to the proposed action:

l (a)

Statement of the specific objectives that the proposed action is

-designed to achieve:

(b)-

General description of the activity that would be required by the licensee or-applicant in order to complete the action:

(c)

Potential change in-the risk to the public from the accidental release of radioactive material:

i (d)

Potential impact on radiological exposure of facility employees and other onsite workers:

(e)

Installation and continuing costs associated with the action, i

including the cost of facility dcwntime or the cost of construction dela;:

(f)

The potential safety im uct of changes in plant or operational complexity, including t1e 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 resources:

(h)

The potential impact of differences in facility type, design, or age on the relevancy and practicality of the proposed action:

Whether the proposed action is interim or final, and if interim, (i) 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 regard:

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7-1.

The proposed priority or schedule.

2.

A summary of the current backlog of existing requirements 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 appropriate with regard to priority, schedule, or cumulative impact.

-comment?ple, could implementation be delayed pending public For exam This item is not applicable to this proposed staff action.

This ger,eric letter only requests existing information and is not a backfit.

(viii)-

For each backfit analyzed pursuant to 10 CFR 50.109(a)(2) (i.e., not

~

adequate protection backfits and not compliance backfits). the proposing Office Director's determination, together with the rationale for the determination based on the consideration of paragraphs (1) and (vii) above, that:

(a) There is a substantial increase in the overall protection of 4 -

)ublic health and safety or the common defense and security to

>e 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.

This item is not applicable to this proposed staff action.

(1x)

For adequate protection or compliance backfits evaluated 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.

1-(b) in addition, for actions that were immediately effective (and therefore issued without prior CRGR review as 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.

This item is not' applicable to this proposed staff action.

(x)

For each evaluation conducted for proposed relaxations or decreases in current requirements or staff positions, the proposing Office -

a.-

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8-l

-Director's determination, together with the rationale for the determination based on the considerations or paragraphs (1) through (vii) above, that:

(a) The public health and safety and the common defense and security would be adequately protected if the proposed reduction in requirements or positions were implemented, and (b) The cost savings attributed to the action would be substantial enough to justify taking the action.

This item is not applicable to this proposed staff action.

(xe) 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 de:cribes the need for the information in terms of potential safety benefit.

(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 impose new requirements on the licensee, other than for the requested information.

The requirement to submit information contained in this generic letter is subject to exception, as discussed in Section Ill.A of the CRGR Charter.

Through 10 CFR 50.54(f) this proposed generic letter requires requests addressees to submit. in writing, the results and certain assum)tions of the current design-basis analysis used to determine the fPSH available for their emergency core cooling and containment heat removal system pumps.

In particular. the generic letter requests addressees to submit details of the containment pressure

-analysis used to determine the containment overpressure. if overpressure is credited in the licensee's NPSH calculation.

The safety benefit of receiving-the requested information is that the staff will be able-to verify tnat the current licensing basis of the plant _ relative to the NPSH available to the subject pumps. meets the design requirements of the pumps under all design-basis accident

. conditions.

The-staff will also be~able to compare the responses in order to further investigate situations at particular plants for which NPSH conditions for certain of the su) ject pumps could

' represent a potential safecy hazard to nuclear power plant operation.

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9 The staf f will begin to review the recuired information as soon as responses begin to be received from acdressees.

The information request regn n eent does not impose new requirements on licensees, other than preparing the information ident1fied in the generic letter.

(xil)

An assessment of how the proposed action relates to the Commission's Safety Goal Policy Statement.

The proposed generic letter constitutes a request for information only

_