ML20008C700

Forwards Documents Associated W/Gl 98-02, Loss of Reactor Coolant Inventory & Associated Potential for Loss of Emergency Mitigation Functions While in Shutdown Condition

ML20008C700
Person / Time
Issue date:	07/02/1998
From:	Shapaker J NRC (Affiliation Not Assigned)
To:	NRC
References
GL-98-02, GL-98-2, TAC-M92635, NUDOCS 9807060045
Download: ML20008C700 (1)

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UNITED STATES i

y NUCLEAR REGULATORY COMMISSION 2

WASHINGTON. D.C. 30615 0001

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July 2. 1998 MEMORANDUM TO:

Document Processing Services Section Records Management Branch Information Management Division j

Office of the Chief Information Officer FROM:

James W. Shapaker Events Assessment -

eneric Commu cations Branch Division of Reactor rogram Management Office of Nuclear Reactor Regulation

SUBJECT:

DOCUMENTS ASSOCIATED WITH NRC GENERIC LETTER 98-02. LOSS OF REACTOR COOLANT INVENTORY AND ASSOCIATED POTENTIAL FOR LOSS 0F EMERGENCV MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION (TAC NO. M92635)

The Reactor Systems Branch (SRXB) in the Division of Systems Safety and Analysis (DSSA) prepared the subject generic letter, which was issued on May 28, 1998, and given accession number 1805050197. There is material related to the subject generic letter that should be placed in the NRC Public q

Document Room and made available to the public. Therefore, by copy of this memorandum. I am providing the following documents to the NRC Public Document Room:

(1) a copy u. the published version of the subject generic letter.

(2) a copy of the information paper (SECY-98-093) that was sent to the Commission. (3) a copy of each letter received in response to the notice of opportunity for public comment on the proposed generic letter that was N

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published in the federal Register on February 14. 1997. (4) the resolution of public comments, and (5) a copy of the review package NRC staff submitted to the Committee to Review Generic Requirements (CRGR).

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 appropriate NUDOCS entries to reflect the fact that the documents identified herein are related to Generic Letter 98-02.

Attachmonts:

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9807060045 9807029 8de/sht CF SUBJ l

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NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 May 28,1998 i

i NRC GENERIC LETTER 98-02: LOSS OF REACTOR COOLANT INVENTORY AND ASSOCIATED POTENTIAL FOR LOSS OF EMERGENCY MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION l

Addressees All holders of operating licenses for pressurized-water reactors (PWRs), except those who have i

permanently ceased operatons, and have certified that fuel has been permanently removed j

from the reactor vessel.

Purpose f

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The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to request that addressees (1) assess the susceptibility of their residual heat removal (RHR) and emergency l

core cooling (ECC) systems to common-cause failure as a result of reactor coolant system (RCS) draindown while in a shutdown condition, and (2) submit certein information, pursuant to

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Section 50.54(f) of Title 10 of the Code ofFederalRegulations (10 CFR 50.54(f)), conceming their findings regarding potential pathways for inadvertent RCS drain-down and the suitability of surveillance, maintenance, modification and operating practices and procedures regarding configuration control during reactor shutdown cooling. The requested information will enable NRC staff to verify whether addressees comply with NRC regulatory requirements and conform l

with current licensing bases for their facilities, with regard to prescribing and accomplishing activities affecting quality per Criterion V of Appendix B to 10 CFR Part 50. The staff is specifically concemed about addressees' controls over the conduct of activities during hot j

shutdown conditions that may affect the safety-related functions of the RHR system and the l

ECCS, for example, the methods used to verify valve position, the controls in place to assure

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compliance with plant surveillance, maintenance, modification and operating procedures, and j

the adequacy of operator training for such activities.

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Discussiori The NRC issued information Notice (IN) 95-03, " Loss of Reactor Coolant inventory and Potential Loss of Emergency Mitigation Functions While in a Shutdown Condition," on January 12,1995, to alert addressees to an incident at the Wolf Creek plant involving the loss

. of reactor coolant inventory while the reactor was in a hot shutdown condition. In that event, 4

operators were attempting to reborcte RHR train B, while at the same time maintenance personnel were repacking an RHR train A-to-train B crossover isolation valve. Train B is i

reborated by recirculating water through a loop that contains the RHR system piping, the i

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GL 98-02 MIy 28,1998 Page 2 of 5 j

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refueling water storage tank (RWST), a containment spray pump, a manual RWST isolation valve, and an RHR system crossover line. When the RWST isolation valve was opened for the reboration process and the train A-to-train B crossover isolation valve was opened for stroke i

testing, a drain-down path was inadvertently created from the RCS to the RWST.

At Wolf Creek, all RHR and ECC system pump suction lines are tied into a common suction i

header. When the draindown event occurred, hot RCS water was introduced into this common suction header between the RWST and the RHR and ECC system pumps. This hot water flashed to steam, resulting in a steam / water mixture in the header. Had an ECCS actuation occurred, this mixture would have been introduced into the suction of the ECCS pumps. If operators had not been able to terminate the event, the hot water in the RWST suction piping might have led to stum blnding, which could have adversely affected the pumps in both ECCS l

trains. In addition, water flashing to steam in the header and the RWST could have caused serious mechanical damage to the RHR piping and the RWST as a result of water hammer.

I Finally, steaming through the RWST establishes a containment bypass path.

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The licensee estimated (using actual plant conditions) that for an unmitigated event, the reactor vessel water level could have drained to the bottom of the hot leg within 5 minutes and, as a consequence, RHR purao A would have lost suction, cavitated, and failed. Shortly thereafter, the common ECCS suction header could have reached a 90-percent steam / water ratio. The i

licensee also estimated that continued boil-off could have caused the pressure vessel water level to drop to the point of core uncovery in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Events of this nature are considered particularly significant because they can result in loss of l

emergency core cooling capability and involve the potential for containment bypass. On March 25,1996, the staff issued a supplement to IN 95-03 that further analyzed the event. The NRC has also issued a number of other communications describing events at reactor facilities j

involving inadvertent loss of reactor coolant inventory while the reactor was in a shutdown I

condition. The Office for Analysis and Evaluation of Operational Data (AEOD) published AEOD/E704," Discharge of Primary Coolant Outside of Containment at PWRs While on RHR Cooling," in March 1987, which documented six events involving RCS backflow into the RWST.

In Generic Letter 88-17. " Loss of Decay Heat Removal (DHR) 10 CFR 50.54(f)," dated October 17,1988, the NRC requested several actions to address loss-of-DHR events that occurred while reactors were in a shutdown condition. In IN 9142," Plant Outage Events Involving Poor Coordination Between Operations and Maintenance Personnel During Valve Testing and Manipulations," dated June 27,1991, the NRC discussed inadvertent loss-of-inventory events. The AEOD report " Reactor Coolant System Blowdown at Wolf Creek on September 17,1994," (AEOD/S95-01), dated March 1995, noted 19 events in which RCS water was transferred to the RWST. These events were primarily caused by personnel errors, poor i

coordinaten between operations and maintenance personnel, and inadequate prncedures assocated with the operation of the RHR system in the shutdown cooling mode. The personnel errors were primarily caused by inattention or lack of training; while the procaf: ural deficiencies were related to omissons or lack of specificity in sequential valve operatens when conducting i

tests on the RHR system. On the basis of this history and the potr..itial for containment bypass, the staff has concluded that additional information is required to confirm the adequacy of existing configuration control, operating practices, and training for assuring the safety fr capability of the RHR and ECC systems.

GL 98-02 M!y 28,1998 Page 3 of 5 Required informaton Within 180 days of the date o'this generic letter, addressees are required to perform the following: (1) an assessment of whether your emergency core cooling systems include certain design features, such as a common pump suction header, which can render the systems susceptible to common-cause failure as a result of events similar to the Wolf Creek RCS drain-down event of September 17,1994; and if this susceptibility is found, (2) prepare, with consider-ation of plant-specific design attributes, a description of the features of your Appendix B quality assurance program (for example, the methods used to verify valve position, the controls in place to assure compliance with plant surveillance, maintenance, modification and operating procedures, and the adequacy of operator training for such activities) that provide assurance that the safety-related functons of the RHR system and ECCS will not be adversely affected by activebes conducted at hot shutdown (such as occurred at Wolf Creek). Addressees may limit their attention to those surveillance, maintenance, modification and operational activities at hot shutdown during which it is feasible to divert RCS fluid to the RWST, resulting in simultaneous drain-down of the RCS and voiding in the suction header for the RHR and ECC system pumps.

Addrerssees may further limit their response to the consideration of potential configurations and conditions that involve flow paths with pipe diameters equal to or greater than 2 inches. If the assessment performed in response to part (1) of the above requested information does not reveal that a susceptibility exists, then no submittal is necessary.

If the assessment performed in response to pa't (1) of the above required information reveals r

that the susceptibility exists, then the result of the asser,sment shall be submitted in writing, pursuant to 10 CFR 50.54(f) and 10 CFR 50.4, to the U.S. Nuclear Regulatory Commission, A1TN: Document Control Desk, Washington, D.C. 20555-0001, signed under oath or

. affirmation under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, with a copy to the appropriate regional administrator and the appropriate NRC resident inspector. The response to part (2) of the above information request need not be submitted to the NRC. However, responses to parts (1) and (2) of the required information shall be kept h a retrievable licensee system that NRC can verify on an as-needed or sample basis.

e Backfit Dscussion This generic letter only requests information from the addressees under the provisions of Secten 182a of the Atomic Energy Act of 1954, as amendsd, and 10 CFR 50.54(f), to verify addressee compliance with the Commission's regulations and conformance with the current licensing-basis of their respective facilities relative to the safety-related functions of the RHR and ECC systems, and the requirements of Appendix B to 10 CFR Part 50. With respect to Appendix E % 10 CFR Part 50, the requested information will enable the NRC staff to determine whether adequate control ls being exercised over surveillance, maivanance, modification and operational activebes conducted at hot shutdown which can adversely affect the safety felated functions of the RHR and ECC systems. No backfit is either intended or approved in the context of issuance of this generic letter. Therefore, the staff has not performed a backfit analpis.

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GL 98-02 Mxy 28,1998 Page 4 of 5 5

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i Quality Assurance Criterion V of Appendix B to 10 CFR Part 50 requires that " activities affecting l

quality shall be prescribed by documented instructions, procedures, or drawings of a type l

appropriate to the circumstances and shall be accomplished in accordance with these i

instructxms, procedures, or drawings." Furthermore, licensees' technical specifications include requirements to establish, implement, and maintain written administrative procedures to address startup, operaton, and shutdown of a shutdown cooling system. Maintenance and i

testing activities at Wolf Creek during hot shutdown were carried out contrary to documented procedures and th a technical specifications, resulting in RCS drain-down and the potential for common-cause fa lure of the RHR and ECC system pumps, which muld have compromised the ability of the RHR and ECC systems to fulfill their safety functions. Furthermore, the staff has i

determined that senitar loss-of-coolant events while on RHR cooling have occurred at over 19 plants. These even'. were due to the failure on the part of licensees to either establish e

adequate procedures or follow procedures and applicable technical specifications. Both of these conditions involve non-compliance with the requirements of Criterion V of Appendix B to 10 CFR Part 50, and, therefore, non-compliance with the current licensing basis for a facility.

Since, a relatively large number of the operating PWRs have experienced similar events, the staff believes that additional information is required to confirm the adequacy of existing configuration control practices, operating practices, and training for assuring the safety function capability of the RHR and ECC systems. In acmrdance with the provisions of 10 CFR 50.54(f),

an approved evaluation of the rationale for the information request contained herein is not a prerequisite to issuance of the generic letter because the information being requested is needed by the NRC staff to verify addressee compliance with the current licensing bases of their respective facilities.

FederalRegister Notoficaton A notice of opportunity for public comment was published in the FederaIRegister (62 FR 7075) on February 14,19g7. Comments were received from four nuclear utility companies, the Nuclear Energy institute, and the Nuclear Utility Backfitting and Reform Group. The staff's evaluation of the comments is available from the NRC Public Document Room. The generic letter has been appropnately revised to reflect the comments received.

Paperwork Reduchon Act Statement This Generic Letter contains information collections that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collechons were approved by the Office of Management and Budget, approval number 31500011, which expires -

September 30,2000.

The public reporting burden for this mandatory information collection is estimated to average 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> per response, including the time for reviewing instructions, searching existing data i

sources, gathering and maintaining the data needed, and completing and reviewing the information collection. The U.S. Nuclear Regulatory Commission is seeking public comment on the potentialimpact of the information collections contained in the generic letter and on the following issues:

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GL 98-02 M:y 28,1998 Page 5 of 5 1.

Is the proposed information collection necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility?

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Is the estimate of burden accurate?

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Is there a way to enhance the quality, utility, and clarity of the information to be collected?

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

Send comments on any aspect of this information collection, including suggestions for reducing the burden, to the Information and Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Intemet electronic mail at BJS1@NRC. GOV; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011), Office of Management and Budget, Washington, DC 20503.

Public Protection Notification if an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

If you have any questions about this matter, please contact the technical contact listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.

Original signed by Jack W. Roe, Acting Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contact: M. M. Razzaque, NRR 301-415-2882 E-mail: mmr1@nrc. gov Lead Project Manager: Kristine Thomas, NRR 301-415-1362 E-mail: kmt@nrc. gov

Attachment:

List of Recently issued NRC Generic Letters DOCUMENT NAME: S:\\DRPM,_SEC\\98-02.GL Ta receive a copy of this document, indicate in the box C= Copy w/o attachment / enclosure E= Copy with attachment / enclosure N = No copy OFFICIAL RECORD COPY l OFFICE SXRB*

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(A)D:DRPM INAME MRazzaque JGoldberg JStolz goe p

5/27/98 3/16/98 5/27/98 N8 lDATE

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Attachment GL 93-02 May 28,:1998 i

Page 1 of 1 I

LIST OF RECENTLY ISSUED GENERIC LETTERS i

GENERIC '

DATE OF i

LETTER -

SUBJECT ISSUANCE ISSUED TO 98-01

' Year 2000 Readiness of 05/12/98 All holders of OLS for i

of Computer Systems at nuclear power plants, j

Nuclear Power Plants except those who have permanently ceased l

operations and have l

certifed that fuel has been permanently removed from the reactor vessel 07-06 Degradation of Steam 12/30/97 All holders of OLS for Generator intemals pressurized-water reactors, except those who have permanently ceased operations and have certifed that fuelhas been perman-ently removed from the j

reactor vessel 97-05 Stum Generator Tube 12/17/97 All holders of OLs for l

Inspection Techniques pressurized-water reactors, i

except those who have -

permanently ceased operations and have certifed that fuel has been perman-ently removed froin the reactor vessel 96-06, Assurance of Equipment 11/13/97 All holders of OLs for nuclear Sup.1 Operability and Containment power reactors except those integrity During Design-Basis who have permanently Accident Conditions ceased operations and have certifed that fuel has been permanently removed from the reactorvessel

. OP = Operating L' ense c

CP = Construction Permit NPR = Nuclear Power Reactors

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POLICY ISSUE Aoril 30.1998 (lnformatIOn)

SECY-98-093 EQB:

The Commissioners FROM L. Joseph Callan Executive Director for Operations

SUBJECT:

PROPOSED NRC GENERIC LETTER 98-xx, LOSS OF REACTOR COOL.NT INVENTORY AND ASSOCIATED POTENTIAL FOR LOSS OF EMERGEN ;Y MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION l

PURPOSE:

To inform the Commission of the staffs intent to issue the subject generic letter. In the generic letter, the staff asks the licensees of all pressurized-wa;er reactors to make available to the NRC certain information [ pursuant to >ection 50.54(f) of Title 10 of the Code of FederalRegulations e

(10 CFR 50.54(f))] regarding the subject matter of this generic letter. This information wi!! enable NRC staff to verify whether addressees comply with NRC regulatory requirements and conform with current licensing bases for their facilities, including the establishment, and conduct of activities affecting quality per Criterion V of Appendix B to 10 CFR Part 50.

A copy of the proposed generic letter is attached (Attachment 1).

DISCUSSION:

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The staff issued Information Notice (IN) 95-C3, " Loss of Reactor Coolant inventory and Potential Loss of Emergency Mitigatic Functions While in a Sr:Jtdown Condition," on January 12,1995, to alert addressees to an incident at the Wolf Creek nuclear power plant ir,volving the loss of reactor coolant inventory while the reactor was in a shutdown condition. On March 25,1996, the staff issued a supplement to IN 95-03 that further analyzed the event and provided additional i

insights. These insights also heightened awareness of the safety significance of similar events.

The draindown event at Wolf Creek represents a shutdown vulnerability that was not recognized earlier. Events of this nature are considered particularly safety significant because loss of coolant can result in a loss of emergency core cooling system capability, and also involves the l

potential for containment bypass. Another important spect of this event is the short time available to the operators for taking corrective action.

4 CONTACT: M. Razzaque,SRXB/DSSA 1

415-2882 TO BE MADE PUBLICLY AVAILABLE AFTER GL IS ISSUED

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The Commissioners 2

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The staff proposes to issue this generic letter to request that addressees determine if their l

emergency core cooling systems (ECCSs) are susceptible to common-cause failure as a result I

of events similar to the Wolf Creek reactor coolant system draindown event of September 17, 1994. If found susceptible, the generic letter requests that information regarding the prevention

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of such events be made available to NRC. This generic letter was endorsed by the Committee 1

to Review Generic Requirements (CRGR) ouring its meeting (Number 291) on September 11, 1996. The staff submitted it to the Commission on November 6,1996 (SECY-96-231). A staff requirements memorandum issued by the Commission on January 22,1997, directed the staff to i

allow the public at least 30 days to comment before the generic letter was issued. As a result, a nobce of oppoitunity for public comment was published in the Federa/ Register on February 14, 1997. In response to the substantial public comments received, the staff revised the genenc letter and prepared responses to the comments (Attachment 2).

In the attached proposed final generic letter,'the staff requests that licensees (1) perform an assessment of whether their ECCSs include certain design features, such as a common ECCS pump suction header, which can render the ECCS suscephth to commorsause failure as a result of events similar to the Wolf Creek reactor coolant system draindown event, and if this suscephbility is found, (2) prepare, with consideration of plant-specific design attnbutes, a desenpbon of the features of their Appendix B quality assurance program that provide assurance that the safety-related funchons of the residual heat removal system and ECCS will not be adversely affected by activsbos conducted at hot shudown (such as occurred at Wolf Creek). If -

the assessment performed in response to part (1) of the above requested informahon reveals that the snacal*Niity exists, then the result of the manaamment must be submitted to NRC. The response to part (2) of the above aformabon request need not be sabmitted to NRC, but must be kept in a retnewable licensee system that NRC can verify on an as-reeded or sample bases The staff will prepare guidance for the inspectors who will perform these venficahons within the currently available resources The CRGR reviewed this revised generic letter during its meeting (Number 314) on January 30, 1998, and the staff has ir.ri,ii,cC2 3 the comments made by the CRGR at that meeting. The CRGR has endorsed the ixc-;-:::$ final genene letter without formal revow. The Advoory Committee on Reactor Safeguards (ACRS) ret iewed this revised genene letter duri T,is 446th meeting on November 6,1997 An ACRS letter raport,.iated November 13,1997, recommended that the proposed final generic letter be promptly issued. The Offica c.

General Counsel has reviewed this generic letter and has no legal obpechons to its coa r.a The staff intends to issue this generic letter approsimately 5 workng days after the date of this afomwbon paper.

//1 Callan a mooi for Operations Attachments:

1. Proposed Generic Letter," Loss of Reactor Coolant inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition"

2. Public Comment Resolution and Staff Response.

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EAR CE E ON March 17,1997 Chief, Rules Review and Directives Branch U. S. Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, D. C. 20555-0001

SUBJECT:

Industry Comments on Proposed NRC Generic Letter 97-XX -

" Loss of Reactor Coolant Inventory and Associated Potential for Losa of Emergency Mitigation Functions While in a Shutdown Condition" (62 Fed. Reg. 7075 - February 14,1997)

Oooortunity for Public Comments The Nuclear Energy Institute 1 offers the following comments relative to a Federal Register notice which solicited public comments on a proposed NRC Generic Letter (GL) requesting licensees to assess the susceptibility of their emergency core cooling system (ECCS) to common-cause failure as a result of reactor coolant system (RCS) drain down while in a shutdown condition.

The proposed GL addresses concerns resulting from the Wolf Creek RCS drain-down event af September 17,1994 (NRC Information Notices 95-03 and j

95-03, Supplement 1). Licensees are regt.ested to assess the susceptibility of their ECCS to common-cause failure as a result of RCS drain-down while in a shutdown condition. Licensees are also requested to submit certain information gursuant to 10 CFR 50.54(f) concerning their findings regarding potential pathways for inadvertent RCS drain-down and the suitability of configuration control and operating practices during reactor shutdown cooling. NRC states it will review this information to verify that licensees are in compliance with General Design Criteria (GDC) 34 (residual heat removal) and GDC 35 (emergency core cooling) of Appendix A to 10 CFR Part 50.

j 3 NEI is the organization responsible for establishing unified nuclear industry policy on matters affceting the nuclear energy industry, including regulatory aspects of generic operational and technical issues. NEl men.bers include all utilities 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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Page 2 Plants are designed and licensed for compliance to GDCs 34 and 35. Suitable consideration of redundancy, single failures, and power supplies ensures reliable j

operation of the ECCS and residual heat removal system (RHRS). In practice, 1

the GDCs and Technical Specifications result in requirements for all ECCSs and RHRSs to be fully operable and single-failure proof during power operation and hot shutdown conditions.

During cold shutdown and refueling conditions, the Technical Specification requirements for ECCSs do not require automatic actuation or that the systems be single-failure proof. In cold shutdown and refueling conditions, design considerations alone cannot preclude all permutation = of valve lineups that could lead to RCS drain-down events. Administrative and work controls must be relied upon for this function. Most licensees utilize a " protected train concept" i

in conjunction with controls over emergent work activities. In addition, the outage plan is configured and controlled to assure the performance of key safety functions. Plants are committed through 10 CFR 50 Appendix B and their Technical Specifications to establish and follow proccdures reflecting their work controls, and are subject to violation if they do not.

Industry recognizes the safety significance of the Wolf Creek event. However, we question the need for a GL as a means to further emphasize the importance of rigorous administrative controls to preclude creation of potential RCS drain-down paths due to simultaneous valve manipulations during outage conditions.

We believe this has been accomplished through issuance of NUMARC 91-06,

" Industry Guideline to Assess Shutdown Management," and industry reviews of the Information Notices resulting from the Wolf Creek event. Further, we believe the action requests contained in the proposed GL are unnecessary given the work controls discussed above.

The proposed GL discusses a potential common-cause failure mode involvmg the ECCS pump suction header. Some PWR plant designs do not incorporate a common header for the ECCS pump suction and/or do not allow the RCS recirculated fluid to enter the shutdown cooling system between the borated water storage tank outlet and pump suction. These plants are not susceptible to the common-cause steam binding event that occurred at Wolf Creek.

Other plant designs incorporate a common header for the ECCS pump suction and, in the case of an RCS drain-down event, the ECCS pumps taking suction from this header could fail due to common cause (steam binding), as the GL notes for the Wolf Creek design. The proposed GL requests actions based on the

" susceptibility" to this failure mode. Susceptibility to this common-cause failure mode is predicated upon errors of commission (valve manipulations) that were

U. S. Nuclear Regulatory Commission March 17,1997 Page 3 violations of administrative controls. Practical design features cannot be established to preclude this or other situations that could be postulated on j

the basis of errors of commission during shutdown conditions. We agree that existence of this particular failure mode places further emphasis on the need for proper administrative controls during shutdown. This has been accomplished through implementation of the industry initiative associated with NUMARC 91-06 and further emphasized through industry awareriess of the Wolf Creek event.

With regard to the specific action requests of the proposed GL, we note the following:

Action request: Licensees are requested to determine whether their ECCSs are susceptible to common-cause failure, e.g., as a result ofevents similar to the Wolf Creek RCS drain doum event.

Comment:. Some (but not all) PWR plant designs utilize a common ECCS suction header and are potentially susceptible to this type of failure mode i

given valve misoperations during shutdown. PWR plants are familiar with the sequence of events described in the Information Notices and the AEOD Special Report on the Wolf Creek event.

j Action request: IfECCSs are found to be susceptible to common-cause failure, licensees are expected to take corrective action, as appropriate, in accordanct with the requirements ofSection XVIofAppendix B to 10 CFR i

Part 50, to ensure compliance with NRC regulatory and licensing requirements.

Comment: Plants have already demonstrated and been licensed on the basis of compliance to GDCs 34 and 35, and design features alone cannot preclude this,;r similar events. Corrective actions have been taken in the form of establishing appropriate work controls to preclude valve misoperations during shutdown conditions Action request: If the RCS is found to be susceptible to drain-down events, j

describe each potential drain-down flow path (include piping sizes, identify flow path valves and their normal positions, and identify valve interlocks

- and provisivit s for valve position indication in the control room), describe potential valve testing manipulations or uses, and describe any administrative controls that are intended to be used to control valve manipulations to preclude RCS drain-down events.

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U. S. Nuclear Regulatory Commission l

March 17,1997 Page 4 Comment This action request represents a major burden on licensees that is neither practical nor necessary. Plant procedures are in place to j

effect administrative controls addressing normal shutdown and testing valve alignments. Categorization and description of the permutations of valve lineups and potential misoperatior.nat could contribute to RCS drain-down events would entail an exhaustive efrort with little practical significance. It is certainly important that each outage operation involving these valves be carefully reviewed with regard to drain-down potential at the time that the outage work is planned or executed. However, to i

sttempt to accomplish this in advance for all possible combinations is simply not practical. The number of combinations is so high that it would be difficult to guarantee an all-inclusive effort. It is fundamentally more prudent to review these valve lineups and outage operations on a case-by-case basis where the scope can be constrained to practical dimensions, and greater assurance provided of an intensive review.

1 With regard to the backfit discussion of the proposed GL offer the following:

If the intent is to interpret 10 CFR 50.46 and the relevant GDCs to suggest that design features must be in place to preclude these types of events during shutdown operations, then we believe the proposed GL represents a significant i

backfit requiring a regulatory analysis. Further, if the intent is for licensees to j

investigate and categorize all permutations of valve operations that could lead to this type of event, we believe this represents a significant revision to current i

practices for regulation relative to outage conditions and should be subject to i

regulatory analysis.

We appreciate the opportunity to provide comments on this proposed Genene i

Letter. Ifyou *.. ave any questions regarding these comments, please contact Biff Bradley at (202)-739-8083.

Sincerely, hL G

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RC-97-0055 Chief, Rules Review and Directives Branch U. S. Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, D. C. 20555-0001 Sir:

Subject:

VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50/395 OPERATING LICENSE NO. hTF-12 COMMENTS ON THE PROPOSED GENERIC IEITER ON LOSS OF REACTOR COOLANT INVENTORY AND ASSOCIATED POTENTIAL FOR LOSS OF EMERGENCY MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION South Carolina Electric & Gas Company (SCE&G) has reviewed the proposed generic letter and has the following comments.

The proposed generic letter ignores licensee resiews and any action taken regarding Information Notice 95-03 and Generic Issue 105, " Interfacing System LOCA in Light Water Reactors." SCE&G recommends the NRC Resident Inspectors appraise the operating experience evaluation of the Information Notice and any utility review / action taken regarding the Generic Issue, instead ofissuing a generic letter. Data collected by this inspection could be added to the next Resident inspector's monthly report. These actions would quickly provide the desired information within the time-frame cf the mitial generic letter response.

It is SCE&G's belief that the information found in these reviews will resolve any NRC concern.

Alternatively, the NRC should provide licensees at least 180 days or until the next refueling outage to respond to the Generic Iet".cr.

If you have any questions, please contact Chris Crowley at (803) 345-4409.

VeryTruly Yo W

Gary kflor cac c: J. L. Skolds W. F. Conway R. R. Mahan R. J. White L. A. Reyes A. R. Johnson NRC Resident Inspector NSRC RTS (MSP 970005)

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..G L Tennessee Valley Authority.1101 Market Street, Chattanooga, Tennessee 37402-2601 March 14, 1997 Chief, Rules Review and Directives Branch U.S.

Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, D.C.

20555-0001 Gentlemen:

INDUSTRY COMMENT ON PROPOSED NRC GENERIC LETTER (GL) 97-XX,

" LOSS OF REACTOR COOLANT INVENTORY AND ASSOCIATED POTENTIAL FOR LOSS OF EMERGENCY MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION" TVA offers the following comments relative to Federal Recister Notice (62 FRN7075, February 14, 1997) which solicited public comments on a proposed NRC GL requesting licensees to assess the susceptibility of their emergency core cooling systems (ECCSs) to common cause failur' as a result of a reactor coolant system drain-down while in a shutdown condition.

The proposed GL requests that licensees determine whether their ECCSs are susceptible to common-cause failure as a result of events similar to the Wolf Creek reactor coolant system (RCS) drain-down event of September 17, 1994.

If ECCSs are found to be susceptible to such common-cause failure, addressees are expected to take corrective action as appropriate in accordance with the requirements stated in Section XVI of Appendix B to 10 CFR Part 50, to ensure compliance with the regulatory guidance provided in general design criteria (GDC) 34 and 35.

Comment:

This action request represents a major burden on licensees that is neither practical nor necessary.

Categorization and description of the permutations of valve lineups and potentia} misoperations that could contribute to RCS drain-down events would entail an exhaustive effort with little practical significance.

It is certainly important that each outage operation involving these valves be kVW.

s Chief, Rules Review and Directives Branch Page'2 March 14, 1997 carefully reviewed with regard to drain-down potential at the time that the outage work is planned.

However, to attempt to accomplish this in advance for all possible combinations is simply not practical.

The number of combinations is so large that it would be difficult to guarantee an inclusive effort.

It is fundamentally more prudent to review these valve lineups and outage operations on a case-by-case basis, where the scope can be constrained to practical dimensions, and greater assurance provided of an intensive review.

In addition, suitable consideration of redundancy, single failures, and power supplies ensures reliable operation of the ECCSs and residual heat removal system, and Technical Specifications preclude automatic ECCSs actuation during cold shutdown conditions.

Administrative and w.rk controls 1

must be relied upon for evolutions described by this event.

TVA utilizes a " protected train concept" as discussed in NUMARC 91-06, in conjunction with controls over emergent work activities.

The outage plan is configured and l

controlled to ensure the performance of key safety functions.

We recognize the safety significance of the Wolf Creek event, but question the need for a GL as a means to further

~

emphasize the importance of rigorous administrative controls to preclude creation of potential RCS drain-down paths due to simultaneous valve manipulations during outage conditions.

We believe that this has essentially been accomplished through implementation of NUMARC 91-06 and through industry reviews of the Information Notices resulting from the Wolf Creek event.

Further, we'believe the actions requested by the proposed GL are unnecessary, given the work controls discussed above.

With regard to the backfit discussion of the proposed GL, we offer the following:

If the intent is to interpret 10 CFR 50.46 and the relevant GDCs to suggest that design features must be in place to preclude these types of events during shutdown operations, then we believe the proposed GL represents a significant backfit requiring a regulatory analysis.

Further, if the intent is for licensees to investigate and categorize all permutations of valve operations that could lead to this type of event, we believe this represents a significant revision to current practices for regulation relat3ve to outage conditions and should be subject to regulatory analysis.

i I

4 Chief, Rules Review and Directives Branch Page 3-March 14',

1997 1

TVA= appreciates-the opportunity to provide comments on this I

~

proposed GL.

If you have any questions regarding these i

comments, please contact E.

W. Whitaker at (423) 751-6369.

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4 Sin

rely,

,,/'

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/)St/Y <t2kkl i

Raul'R. Baron General Manager Nuclear Assurance and Licensing j

cc:

Mr. Ronald W.

Hernan, Senior Project Manager U.S.

Nuclear Regulatory Commission One White Flint' North 1

11555 Rockville Pike i

Rockville, Maryland 20852

.l Mr. Robert E.

Martin, Senior Project Manager U.S.

Nuclear Regulatory Commission

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one White Flint North 11555 Rockville Pike Rockville, Maryland 20852 Mr. Luis Reyes, Regional Administrator

.U.S.

Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite ' 900 Atlanta, Georgia'30323 Mr. M.

C.

Thadani, Project Manager U.S.

Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike I

Rockville, Maryland 20852 Mr.

J.

F. Williams, Project Manager U.S.

Nuclear Regulatory Commission one White Flint Nortu 13555 Rockville Pike Rt ;kville, Maryland 20852 I'

cc:

Continued on page 4

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NRCLResident Inspector Sequoyah Nuclear Plant 2600 Igou_ Ferry Road j

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Rules Reviewand Directivos Branch Ma!! Stop T4D49

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RE:

Proposed Generie Letter, " Loss of Reactor Coolant Inventcry and Aaseeisted Potential ter Loss of Essergency Mitignalen Fusadons WhDe in a Bhutdown f,ggdglp." d2 Fed. aw 7.39di gm ___,1( t w n ATTN: Chief, Rules Revfew and Directives Branch On February 14,1997, the Nuclear Regulatory Commission ("NRC") issued its above-captioned proposed Generic Letter for public ra=== a Provided below are the comunants of the Nuclear Utility BackStting and Refoem Oroup ("NUBARG").F These mammenen concern the backfhtine impliantions of the proposed Ocasric Imeer.

In the proposed Gensric Letter, the NRC requests licensees to (1) assoas the susospeibility of their samsgancy core cooling systems ("ECCS") e common cause ibhse as a result ofremot:t coolant system ("RCS") drain down while in a shutdown a=M= and (2) submit lan==aden pursuant to 10 C.F.R. i50.54(f) concoming their Sndings regenhag potential pathways forinadvertent RC8 dmn-down and the suitability of con 53uration control and operesing practices during reactor shadows cooling. The 5#h= is to enable the NRC Staff to vert (r whether addressoas coa.piy and confban with NRC regulatory and lloross ip----

~. spectSon!!y verifying the ahquacy of mainemining the sesidual hast removal suisty fbnetion to transfer 5ssion WM doesy heat and other residual heat 60m tbs reactor (Gensual Design Criterica M of Appendix A to Part 50) and the BCC5 % provide ahamdant emergency core cooling wbse required (Genseni Deden Crhodon 35).

The pamoular evast doenGud is the proposed Gensde Lemer concemed asness ant of outage activities retar than system design adequacy. Certain actrvides won ongoing that, when perfbrmed,s 4, resuhod in an open dmin dcwn path Aos the RCS to the rufholing water F

NUBARO is a, consortium of 15 utilities fonned in the sedy 1980s which participated motively in the devd& of the NRC's backfittag nde (10 C.F.R. 650.109) in 1985, and which ha? closely monitored the NRC's application of the rule since tbst time. 7O$/903f2:

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March 17,1997 Page 2 storage tank. De "Backfit Discussion" of the proposed Generic Leuer refers 'o compliance with 10 C.F.R. l50.46, " Acceptance Criteds for Emergency Core Cooling Systs.as for Light Water Nuclear Power Raestors," to jus #y the information toquest to ennnna the Weecy of existing ECCS configutation control and ope =dag practica regrding residual best removal, ne "Backfst Discussion" of the proposed Genedc letter states:

i The actions requestae in this generic lettr!, if required, would be backfits in acconience with NRC ph and are no masary to ensure that addressess are in connpliance with existing NRC rules and rsci-d== WMily,10 CFR 50.46 requires that the ECCS b t desianed to provide adequaes Sow capability to== detain the core temperanne at an ecospeably low value and to remove decay bout for the extended pedod of time raquired by the long lived radiceenGy renalning in the com. b Wolf Creek event has dananstrated that the "7y of ECCG oonfigtmics control and operating,=*s regenhag rusidust best removal can adveroclyimpost ECCS per===r=

and could prevent the ECC3 fhun perfbeeing e

its ashty fbaction ibilowlag events at reactor theilities havolving inadvertant loss of reactor coolant k.

ay while the reactor is shut down. %

siba, this puneric leant is being issued as if the requested actions were cumplianos beskfits under the tenna of10 CFR 50.109(aX4Xi),

l We believe that serie::s requested in the proposed Generic Iseter me insyproprinnely characterized as==l"e exceptions to the backfitting provisions of Bestion 50.109. The i

requirements of Section 50.46 reises to specifle derian fasames of the BCCS, whereas the problem daaa'O=d in the psoposed Gemmic Isiser relaans to "the adequacy of ECC5 confignestion connot and operating practices." The actions requemed by the proposed Osamic Latter more,,4--My relate to conduet and coonlination of andvities while in a shut down condition. By citing Section 50A6 as tbs basis fbr the cesapliance exception to the backfit provisions of 10 C.F.R. j$0.109, tbc propoemd Generic Leaur impth,. that the ECCf must be designed to prevent seek soonarios, when lieeneses geneally rely on adannistrative controls to prevent placing the RCS and ECC8 in such a confignados. Acconting to the NRC's Stuadard Review Plan, the BCCS is designed to rodill "the vessel in a thnely manne hr a lose of. coolant socidsert resulting Aar a spectrusa of postuisand piping beesks within the resseor coolant pressure bounday.'9' Even though opereene actions may result in a poessdal peshway for loss of reesear coolant inventory during shut down a=Maan tbs consequenas me not -- with a pipe hiosk at fh11 power opendens and =ama-d--

to the desip hatues ofthe BCCt1 naqract be the most appropdses conostive astions to address this simassion. Tbs NRC Staff position implied in tbs proposed Osaarle Letter appease to be a new inestproarian of the reguistions in Scotton 50.46 whisk would be subject to the backfhting provisions of 5estica50.109.

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U.5.NuclearRegulacry Commission i

March 17,1997 l

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Licensees have been made aware of the event discussed in the proposed Generic i

Leuer by Efonnation Natioe 95-03 and other generio oesnaraiendons refersnood in the psopossd j

Generic Lacer. Ileanam' aedons taken as a result of these prevums notifications to provost i

inadvertant operssor scions that crases a drain down pushway may be inspeessa by NRC. The

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infunnation sequest is tumoossenly burdensome in scope and may require ibe institudon of studies, i

or other extensive afbrt to asnarete the necessary infonnation to respond. "Ihe proposed Generic l

Lareer would ask licensees, if tbs RCS is found to be susespelle to drain down events, to desen%e

" etch po mmetal drain.down flow path (hslude piping anase, M flow path valves and their e

i nonnal pcehions, and identify valve interlocks and provisions for valve position ladiestion in the j

contro! soom), describe poundd valve testing

• , '"=: or'uses, and describe any j

d=i='=a ive e.mmis that ass imunded to be used to cannon valve manipulatons to pactode Res drainalown events."

we believe that ibis type ofreque,t is unwessansed sad thee the Staf has not shown i

that the burden on licensees is jusified. In the Statement of r==ida=daa for the revision of Basilon 50.54(f), die NRC mmes that "[i]f extensive oftbrt is reasonsbty anticipessd, tbs roguest w!!!

he evskmand to dsesonine whether the beian hnpoemd by tbs infann= nan aquest is justi6ed in view i-of the poesntial ashly significance of ths issus to be addressed.... Ragesses he inhunstion to l

detenedne complianse with amisring fuellity requiremanes... unnally are not made pursuost to l

650.54(f).... The amendment of {$0.54(f) should be read as indicadng a almas conossa on the part i

of the a== nth that exannsive intbansden requ sts be sessihlly sarneinised by seatmanagement c

prior to initiating such regassts. The mamianiaa secognisse that these may be instammes wbsse it e

i is not clear wheear a backfit will fonow an infannation regnest. Thoss eases abould be secolved

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la hver ofanalysis# We believe this and the lansuess of tbs rule isselfindisses the comunission's original latset tbst Session 30.54(f) be used only tr the most sipdficant issues when the cammissles must deementes.msene er na me linense of a asuhrwndd be media.d. supended, or revoked."8' l

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Management controls ibt work activities in shut down opassions, when properly implamamed, provide a reasonable means of reviewing possible velve cosabhutions that could be 1

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'imielwidaamd during speciSc work activities on a ease by. eses heels. Lisensees have basa made swam ofibe in,orianos ofproper adeninimative concels by tbs gunsris a -w-aiandaan merensed in abe proposed Osmaris 14esar. We rosessmand thsa the NRC mot inns the proposed j

Geesris Imiest intil a baskfhting analysis has been compissed, justi$ing the ased for the j

M=amian ad anynewinsepssadens ofte regulations. If tbs Staffbelieves est it has additional i

50 Fed. Reg. 38,112 (Seposeber 20,1985).

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10 C.F.R 650.54(f).

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WD(STON sk STRAWN U. 5. Nuclear Regulsory Comminaion March 17,1997 Page4 M or insights unsibi to licensecs, a secand supplement to Infonastion Notice 95-03 could be issued resher than the proposed Generic Letter.

Sincurs!y, DanielF.Sisager Punisin L. Campbeu CounseltrNuclear UtiL"y Backfhang and Rden Group i

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!O PO Box 1551 All Foyetteville Street Mall k.i L Raleigh NC 27602

-...3 3erial: PE& RAS-97-024 l

1 March 14,1997 Chief, Rules Review and Directives Branch U.S. Nudear Regulatory Commission Mail Stop T-6D-69 Washington, DC 20555-0001

Subject:

Comments on Proposed NRC Generic i etter on Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition (62 FR 7075)

Dear Sir or Madam.

)

This letter conveys Carolina Power & Light Company's (CP&L's) comments on the Proposed Generic Communication regarding " Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition." The proposed letter requests that licensees " determine whether their ECCSs are susceptible to common-cause failure, e.g., as a result of events similar to the Wolf Creek RCS drain-down event of September 17,1994." It further requests that,"If the RCS is found to be susceptible to drain-down events, describe each potential drain-down flow path (include piping sizes, identify flow path valves and their normal positions, and identify valve interlocks and provisions for valve position indication in the control room), describe potential valve testing manipulations or uses, ar.d describe any administrative controls that are intended to be used to control valve manipulations to preclude RCS drain-down events."

CP&L offers the following comments regarding the proposed letter:

Although it appears from the referenced events that the generic letter is specifically concerned with susceptibility of PWR ECCSs to common-cause failure, the scope of the requested response is unclear. It is recommended that the generic letter be clarified as to whether it is intended to address common cause failures of PWR ECCSs or susceptibility of PWRs to drain-down events.

j The event that was the genesis of this proposed generic letter occurred in September e

1994. Subsequently, the NRC issued Information Notice 95-03 and its supplement.

Therefore, potential vulnerabilities identified by this Wolf Creek event may already j

1 t

v d

8 Chief Rules Review and Directives Branch 2

March 14.1997 have been corrected. It is recommended that the NRC assess in this light whether the generic letter is warranted.

The proposed letter requests PWR licensees to evaluate the susceptibility of their ECCSs to common cause failure and implement corrective actions in accordance with 10 CFR 50, Appendix B. Criterion XVI if they are found to be susceptible. CP&L agrees that this request has merit whether based upon the NRC Information Notice or this proposed generic letter. Ilowever, the second part of the letter requests licensees to provide a large volume ofinformation concerning each potential drain-down flow path. CP&L questions if there is any additional safety benefit from providing this large amount of detailed data sinc; any relevant data of this nature will be available on-site following the requested evaluation.

If you have questions regarding this letter, please contact me at (919) 546-6901, or Mike Murdock at (919) 546-3193.

Sincerely.

T.D. Walt Manager, Performance Evaluation

& Regulatory Affairs MLM/

1 cc:

Mr. J. B. Brady, USNRC Resident Inspector - IINP, Unit 1 Mr. B. B. Desai, USNRC Resident Inspector - IIBRSEP, Unit 2 i

Mr. N. B. Le, USNRC Project Manager - liNP, Unit I Ms. B. L. Mozafati, USNRC Project Manager - liBRSEP, Unit 2 i

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1. 1. @ WINSTON & STRAWN i

C 35 WEST WACKER DRIVE 1400 L STREET, N W.

6. RUE DU CIRQUE me mRis. FRANCE CH6CAGO, ILLINOIS 60601 5703 WASHINGTON, D C. 20005 3502 200 MAK AVENUE

43. RUE DU RHONE (202) 371-5700 NEW YORK. NY 10166 4193 1/D4 GEN"VA. SWITZERLAND 1

FACSMILE (202) 371-5950 March 17,1997 y'

j U.S. Nuclear Regulatory Commission 2

i Rules Review and Directives Branch

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Mail Stop T-6D-69 E

e Washington, D.C. 20555-0001 h-j 3

RE:

Proposed Generic Letter," Loss of Reactor Coolant Inventory and As ciath!

Potential for Loss of Emergency Mitigation Functions While in a Sh@down Condition." 62 Fed. Rec. 7.075 (Februarv 14.1997)

ATTN: Chief, Rules Review and Directives Branch On February 14, 1997, the Nuclear Regulatory Commission ("NRC") issued the above-captioned proposed Generic Letter for public comment. Provided below are the comments of the Nuclear Utility Backfitting and Reform Group ("NUBARG").I' These comments concem the backfitting implications of the proposed Generic Letter.

In the proposed Generic Letter, the NRC requests licensees to (1) assess the susceptibility of their emergency core cooling systems ("ECCS") to common-cause failure as a result ofreactor coolant system ("RCS") drain down while :n a shutdown condition, and (2) submit information pursuant to 10 C F.R. {50.54(f) conceming their findings regarding potential pathways forinadvenent RCS dra' Jcwn and the suitability ofconfiguration control and operating practices during reactor shutdom. cooling. The information is to enable the NRC Staff to verify whether addressees comply and conform with NRC regulatory and licence requirements, specifically i

verifying the adequacy of maintaining the residual heat removal safety function to transfer fission product decay heat and other residual heat from the reactor (General Design Criterion 34 of Appendix A to Pan 50) and the ECCS to provide abundant emergency core cooling when required (General Design Criterion 35).

'Ihe panicular event described in the proposed Generic Letter concerned management ofoutage activities rather than system design adequacy. Certain activities were ongoing that, when performed concurrently, resulted in an open drain-down path from the RCS to the refueling water l'

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

NOn DdOd67 3

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WINSTON & STRAWN l

U. S. Nuclear Regulacry Commission March 17,1997 Page 2 j

J storage tank. The "Backfit Discussion" of the proposed Generic Letter refers to compliance with 10 C.F.R. f50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors," to justify the information request to confirm the adequacy of existing ECCS configuration control and operating practices regardine residual heat removal. The "Backfit Discussion" of the proposed Generic Letter states:

The actions requested in this generic letter, if required, would be backfits in accordance with NRC procedures and are necessarv to ensure that addressees are in compliance with existing NRC rules and regulatPns. Specifically,10 CFR 50.46 requires that the ECCS be designed to provide adequate flow capability to maintain j

the core temperatute at an acceptably low value and to remove decay heat for the i

extended period of time required by the long-lived radioactivity remaining in the i

core. The Wolf Creek event has demonstrated that the r (equar r of ECC'1 configuration control and operating practices regarding resi^

1-adversely impact ECCS performance and could prevent th:.CCd iron performit

)

its safety function following events at reactor facildies inve. 4g inadve tent losw i

reactor coolant inventory while the reactor is shut down.

N. t.iis generi-letter is being issued as if the requested actions were compliana

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terms of 10 CFR 50.109(a)(4)(i).

1 We believe that actions requested in the proposed Generic Letter are

'omp tely

)

characterized as compliance exceptions to the backfitting provisns of Section S. The requirements of Section 50.46 relate to specific design fermres of the ECCS, whereas,

4 described in the proposed Generic Letter relates to "the adxuacy of ECCS configuration coe ad operating practices." 'Ihe actions requested by the proposed Generic Letter more appropriatdy relate to conduct U-i coordination of activities while in a shut down condition. By dting Section 50.46 as the basis fo: 4 te compliance exception to the backfit provisions of 10 C.F.R. {50.109, the proposed Generic Letier implies that the ECCS must be designed to prevent such scenarios, when licensees generally rely on administrative controls to prevent placing the RCS and ECCS in such a configuration. According to the NRC's Standard Review Plan, the ECCS is designed to refill "the vessel in a timely manner for a loss-of-coolant accident resulting from a spectrum of postulated piping breaks within the reactor coolant pressure boundary."2' Even though operator actions may result in a potential pathway for loss of reactor coolant inventory during shut down conditions, the consequences are not commensurate with a pipe break at full power operations and modifications to the desigt features of the ECCS may not be the most appropriate corrective actions to address this situation. The NRC Staff position implied in the proposed Generic Letter appears to be a new interpretation of the regulations in Section 50.46 which would be subject to the backfitting provisions of Section 50.109.

1 2'

See NUREG-0800, Section 15.6.5, Revision 2. July 1981.

~

a WINSTON & STILiWN U. S. Nuclear Regulacry Commission March 17,1997 Page 3 Licensees have been made aware of the event discussed in the proposed Generic Letter by Information Notice 95-03 and other generic communications referenced in the proposed Generic Letter. Licensees' actions taken as a result of these previous notifications to prevent inadvertent operator actions that crer.te a drain-down pathway may be inspected by NRC. The information request is unnecessarily burdensome in scope and may require the institution of studies, or other extensive effert to generate the necessary information to respond. The proposed Generic Letter would ask licensees, if the RCS is found to be susceptible to drain down events, to describe "each potential drain-down flow path (include piping sizes, identify flow path valves and their nonnal positions, and identify valve interlocks and provisions for valve position indication in the control room), describe potential valve testing manipulations or uses, and describe any administmtive ecntrols that are intended to be used to control valve manipulations to preclude RCS drain-down events.

We believe that this type of request is un carranted and that the Staff has not shov n

' hat the burden ma licensees is justified. In the Statement of Considerations for the revision of Section aG.!KO, the NRC states that "[i]f extensive effort is reasonably anticipated, the request will be evaluated to determine whether the burden impoced by the information request isjustified 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(0.... The amendment of {50.54(0 should be read ns 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 net clear whether a backfit will follow an information request. Those cases should be resolved in favor of analysis.'T We believe this and the language of the rule itselfindicate the Commission's -

original intent that Section 50.54(0 be used only for the most significant issues when the Commission must detennine whether or not the license of a facility "should 1 e modified, suspended, j

or revoked.'T Management controls for work activities in shut down operations, when properly implemented, provide a reasonable means of reviewing possible valve combinations that could be inadvertently mispositioned during specific work activities on a case-by-case basis. Licensees have been made aware of the importaner ofproper administrative controls by the generic communications referenced in the proposed Geners Letter. We recommend that the NRC not issue the proposed Generic Letter until a backfitting analysis has been completed, justifying the need for the information and any new interpretations of the regulations. If the Staff believes that it has additional F

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

i 10 C.F.R. {50.54(0 l

l MTNSTON & STRAWN U. S. Nuclear Regulaoxy Commission March 17,1997 Page 4 information or insights useful to licensees, a second supplement to Information Notice 95-03 could be issued rather than the proposed Generic Letter.

Sincerely.

Daniel F. Stenger Patricia L Campbell Counsel for Nuclear Utility Backfit:ing and Reform Group l

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l Cuantos 11. Carst

,,..,m Baltirnore Gas and Electne Company

\\' ice President Cahen Cliffs Nuclear Power Plant 1650 Cah en Cliffs Parkw.i>

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March 18,1997 U. S. Nuclear RegWratory Commission Washington, DC 20555 ATTENTION:

Rules Review and Directives Branch

SUBJECT:

Calvert Cliffs Nuclear Power Plant' Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Comments on Proposed Generic Letter: " Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition" The Baltimore Gas and Electric Company is pleased to provide comments on the proposed Generic Letter. We have reviewed and endorse comments submitted by the Nuclear Energy Institute.

This proposed generic letter would: (1) assess the susceptibility of their Emergency Core Cooling System to common-cause fai!u e as a result of Reactor Coolant System drain-down while in a shutdown condition; and (2) submit cenain information concerning their findings regarding potential pathways for inadvenent Reactor Coolant System drain-down, and the suitability of configuration control and operating practices during reactor shutdown cooling.

We feel that the proposed generic letter should not be issued. We feel BGE's existing design provides an adequate basis for assuring that the Emergency Core Cooling System is not subject to common-cause failure as a result of Reactor Coolant System drain-down while in a shutdown condition. The event described by the proposed generic letter does not represent a possibility for common mode failure at Calven Cliffs. Baltimore Gas and Electric Company's design does not allow one shutdown cooling loop aligned to the Reactor Coolant system, with a second loop aligned to the tefueling water storage tank.

Baltimore Gas and Electric Company's design also includes separate suction headers from the refueling water tank to the Emergency Core Cooling System pumps. Calven Cliffs and similarly designed plants should not be subject to the proposed generic letter.

We recommend the Nuclear Regulatory Commission further research the appropriate applicability of this concern prior to issuing a generic letter applicable to all nuclear power plants.

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Rules Review and Directives Branch l

A March 18,1997 Page 2 I

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' Should you have questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours, l

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' Document Control Desk, NRC H. J. Miller, NRC D. A. Brune, Esquire Resident Inspector, NRC j

J. E. Silberg, Esquire R. I. McLean, DNR j

Director, Project Directorate I-1, NRC J. H. Walter, PSC A. W. Dreerick, NRC l

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PUBLIC COMMENT RESOLUTION AND STAFF RESPONSE Substantial comments were received from public/ industry sources as a result of publishing the t

proposed Wolf Creek generic letter (GL) in the Federal Register. These comments have been consolidated by the staff, and are presented below. The originalletters from the public/ industry i

with their comments are attached herewith.

Comment # 1: [SCE&G/W&S/NEl/TVA/CP&t/BGE] The proposed GL ignores licensee reviews and any schon taken regarding previous NRC notifications (IN 95-03 & Supplement, NUMARC

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91-06: " industry Guidelines to Assess Shutdown Management," and Generic issue 105:

j "interfacmg System LOCA in LWR"). Corrective schons have been taken in the form of estab-i lishing appwpr;.te work controls to preclude valve misoperatens during shutdown conditions, j

We recommend the NRC Resident inspectors (RI) appraise the operating exponence evaluation of the IN and other generic communications, and any utility revow/ action taken regarding the Gar:eric issue, instead of issuing a GL Data collected by this inspechon could be added to the next Rl's monthly report. These actions would quicidy provide the desired informabon within the time-frame of the initial GL response. It is our belief that the information found in these reviews will resolve any NRC concem Attematively, the NRC should provide licensees at least 180 days or until the next refueling outage to respond to the GL 1

Some (but not all) PWR plant designs utilize a common ECCS suchon header and are potentially susceptible to this type of failure mode given valve misoperations during shutdown, PWR plants are familar with the sequence of events desenbod in the IN and the AEOD Specal Report on the Wolf Creek event. BGE ciasms that it's plant design does not allow one shutdown cooling loop abgned to the RCS with a second loop aligned to the RWST, and that the design also includes separate suchon headers from the RWST to the ECCS pumps. As a result, these plants provide an adequate basis for assunng that the ECCS is not subject to common-cause fadure in the event of an RCS dram-down while in a shutdown cornhtson. BGE, therefore, ciasms that the plants hovmg such design should not be subject to the proposed GL They recommanded that NRC further research the appropnate applicability of this concem prior to issumg a GL apphcable to all nuclear power plants.

The achon requested by the proposed GL represents major burden on bconsees that is neither prachcal nor necessary. Pl.... procedures are in place to effect admeistrative controls addressmg normal shutdown and testing valve ahgnments. Categontabon and desenphon of the permutebons of valve lineups and potential misopersbons that could contnbute to RCS drandown events would entail an exhaustive effort with little prachcal segruficance it is certainly important that each outage operabon involving these valves be carefully reviewed with regard to drandown potenbal at the time that the outage work is planned or executed. However, to attempt to =+r-;'r. this in advance for all possible combmabons is sempty not prachcal. The number of combinatens is so high that it would be difficult to guarantee an all-inclusrve effort. It is fundamentally more prudent to revow these valve Enoups and outage operations on a case-by-case basis where the scope can be constrained to practical dimensens, and greater assurance provuted of an 'mtenseve review.

The information requested in the proposed GL is unnecessarily burdensome in scope and may require the institution of studies, or other extensive effort to generate the necessary information i

to respond. We believe that this type of request is unwarranted and that the Staff has not shown that the burden on licensees is justified.

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Response if the licensees whose plants are susceptible to Wolf Creek like event, have indeed i

taken corrective actions to protect against the event (as suggested by the comments), then it is not unduly burdensome nor is it unreasonable for NRC to request for this information to be made l

F available in 120 days. Any corrective actions taken are required to be documented and reported

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to appropriate levels of management in accordance with Section XVI of Appendix B to 10 CFR j

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Part 50. This kind of request for information by NRC is not done routinely, but only in special t

circumstances such as the Wolf Creek event, which was the most significant precursor event of 1

1994 with a conditional core damage probability estimated to be 3.0E-3. However, in order to j

further relax the requirements, the GL has been revised such that it requests information at this l

time, and no b&ckT,G,ii will be required. Furthermore, the time allowed to prepare the responses j

has been increased to 180 days. In the revised GL, the staff requests that addressees perform

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l the follomng: (1) an assessment of whether your emergency core cooling systems include certain design features, such as a common pump suction header, whch can render the systems susceptible to common-cause failure as a result of events similar to the Wolf Creek RCS drain-

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dawn event of September 17,1994; and if this susceptibility is found, (2) prepare, with conside. -

i ation of plant-specific desen attributes, a description of the features of your Appendix B quality i

j assurance program (for example, the methods used to verify valve position, the controls in place to assure compliance with plant surveillance, maintenance, modification and operstmg j

procedures, and the adequacy of operator training for such activities) that provide assurance that j

the safety-related functions of the RHR system and ECCS will not be adversely affected by

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activities conducted at hot shutdown (such as occurred at Wolf Creek).

if the assessment performed in response to part (1) of the above required inic,rmation reveals that the suweptibility exists, then the result of the assessment must be submitted to NRC. And, j

if the mamanament does not reveal that a susceptibility exists, then no submittal is necessary.

However, responses to parts (1) and (2) of the required informahon shall be kept in a retrievable l

licensee system that NRC can venfy on an as-needed or sample basis.

i The staff agrees that the scope of the requested information in the proposed GL was too broad, i

l and that a more focused review would be sufficient to address the issue. Based on the Wolf t

Creek exponence, the staff believes that the configurations, condstens and processes dunng 1

shutdown which are most risk-significant must be addressed. Hence, the " Required informaten" g

sechon of the GL has been modified to read: " Addressees may limit their attenbon to those survesitance, maintenance, modification and opershonal activities at hot shutdown dunng which it is feasible to divert RCS fluid to the RWST, resulting in simultaneous drain-down of the RCS and voidmg in th auction header for the RHR and ECC system pumps. Addressees may further limit their response to the considersbon of potential configuratens and condsbons that involve flow

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paths with pipe diameters equal to or greater than 2 inches."

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Comment # 2. [NEl/W&S/TVA/CP&tJBGE] Plants have already demonstrated and been h

leensed on the basis of compliance with GDCs 34 and 35, and design features alone cannot t'

produde this or smier events. If the intent is to interpret 10CFR50.46 and the relevant GDCs to suggest that design features must be in place to preclude these types of events dunng shutdown operatens, then we beheve the proposed GL represents a significant backfit requinng a j

regulatory analysis. Further, if the intent is for imensees to 'mvestigate and categorme all j

permutatens of valve operatens that could lead to this type of event, we believe this represents a significant revision to current F.ctices for regulation relative to outage conditions and should i

l be subject to regulatory analysis.

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The particular event described in the proposed GL concemed management of outage activities rather than system design adequacy. We believe that actions requested in the proposed GL are k

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j inapp'.epriately characterized as compliance exceptions to the backfitting provisions of Section 3

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50.109. The requirements of Section 50.46 relate to specific des:en features of the ECCS, l

whereas the problem described in the proposed GL relates to "the adequacy of ECCS configu-j ration control and operating practices." The actions requested by the proposed GL more appropriately relate to conduct and coordination of activities while in a shutdown condition. By citing Section 50.46 as the basis for the compliance exception to the backfit provisions of

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10CFR50.10g, the proposed GL implies that the ECCS must be desigr:ed to prevent such j

scenarios, when licensees generally rely on administrative controls to prevent placing the RCS i.

and ECCS in such a configuration.

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Acesiding to the NRC's SRP, the ECC3 is designed to refill "the vessel in a timely manner for a l

LOCA resulbng from a spectrum of postulated piping breaks within the reactor coolant pressure boundary"(Sec.15.6.5, NUREG-0800, Rev.2, July '81). Even though operator actions may result in a potential pathway for loss of reactor coolant inventory during shutdown conditsons, the j

consequences ar:

i commensurate with a pipe break at full power opeWons and modifica-t tions to the desig?.,.atures of the ECCS may not be the most apprepr;ste correctrve schons to 1

address this situabon j~

The NRC Staff posebon implied in the proposed GL appears to be a new interpretabon of the j

reguisbons in Sechon 50.46 which would be subject to the backfitting provisions of Sechon i

50.109.

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Response The Wolf Creek-like scenario is credible for some PWR plants that utilize a common j.

ECCS suchon header. The facility may have been designed in accordance with GDC 34 and 35, as the comments ciasm, but if W 'f Creek-like event occurs, there exists the potenbalthat the I

funchons as defined in the tea el specificebons (TS) and GDCs 34 and 35 will be affected for these plants due to common-cause failures of all the RHR and ECCS pumps. The staffis specifically concemed about the quality control of activsbes (for example, the methods used to j

.venty valve possbon, the controls in place to assure comphance with the plant operating proco-dures, and the adequacy of operator traerung for such activebes) conducted dunng hot shutdown i

condsbons affectag the =i:ia:Med funchons of the RHR system and the ECCS, as defined in l

10 CFR Part 50, Appendix A, General Design Critona (GDC) 34 and 35, r=g+f;f. Cnterion V i

L of Appendix B to 10 CFR Part 50 requres that "activettes affecting quality shall be presmbed by i

documented instruchons, procedures, or drawngs of a type appropriate to the circumstances and

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shall be ac ~implished in accordance with these instructions, procedures, or drawings."

l Furthermore, licensees' TS include requuements to establish, implement and maintain written i

administrative procedures to address startup, opershon and shutdown of a shutdown cooling i

system. Mantenance and testing activebes at Wolf Creek during hot shutdown resulted in the i

RCS dram dowr. and the potential for common cause failure of the ECCS pumps, which could s

t have s,,,,,,,vr.i:j the ability of the RHR and ECCS r,:M,e to fulfill the safety furv: tens j

specified in GDC 34 and 35, i=r+T;d;.

~ The GL has been revised to request only informabon from the addressees under the pronsens of 10 CFR 50.54(f). In view of the Wolf Creek draindown event, this information is needed to verify bconsees' comphance with NRC regulatory requirements and current licensing bases for their facehbes as related to the requirements of Criterion V of Appendix B to 10 CFR Part 50, Eg+Mn "; as regards the quality control of activities which can adversely affect the safety-related funebons of the RHR and ECC systems, whose requirements are defined in 10 CFR. Part 50, Appendix A, GDC 34 and 35.

Comment # 3: [CP&L) Although it appears from the referenced events that the generic it.tter is

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l specifically concemed with susceptibility of PWR ECCSs to common-cause failure, the scope of j

the requested response is unclear. It is recomm.ided that the generic letter be Marified as to i

j whether it is intended to address common cause failure of PWR ECCSs or susceptibility of l

j PWRs to drain-down events.

l Resoonse The GL has been clarified in the " Required information" section which states, l

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" Addressees may limit their attention to those surveillance, maintenance, modification and opera-j tional activities at hot shutdown during which it is feasible to divert RCS fluid to the RWST, j

resulting in simultaneous drain-down of the RCS and voiding in the suction header for the RHR I

i and ECC system pumps. Addressees may further limit their response to the consideration of potential configurations and conditions that involve flow paths with pipe diameters equal to or

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greater than 2 inches."

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Co,en, erd # 4: [W&S) Management controls for work activities in shutdown opershons, when

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property implemented, provide a reasonable means of reviewing possible valve combinations j

that could be inadvertently misposiboned during specific work activities on a case-by-case basis.

Licensees have been made aware of the importance of proper administrative controls by the I

genenc communicabons referenced in the proposed GL We recommend that the NRC not issue the proposed GL until a backfitting analysis has been completed, justifying the need for the l

informabon and any new interpretabon of the regulations. If the staff believes that it has addi-tional informahon or insights useful to licensees, a second supplement to IN 95-03 could be 1

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issued rather than the proposed GL l

j Response: The GL requirements i sve been relaxed to request information only, pursuant to 10 CFR 50.54(f), and that backfitting is not required by the GL at this time. Howevor NRC needs the information, as discussed in the GL, to enable NRC staff to venfy whether addressees comply with NRC regulatory requirements and conform with current hcensmg bases for their facilibes, includmg the establishment of, and conduct of activities affecting quahty according to, j

documented procedures, per Criterion V of Appendix B to 10 CFR Part 50.

Comment # 5 [W&S] in the Statement of Considerations for the revision of Sochon 50.54(f), the NRC states that "if extensive effort is reasonably anticipated, the request will be evaluated to

. determme whether the burden imposed by the informaton request is justified in view of the y

potential safety significance of the issue to be ddressed...... Requests for informabon to j

determine c,n ;;er.co with existing facility requirement:..... usually are not made pursuant to L

e 50.54(f).... The amendment o,50.54(f) should be read as indicating a strong concem on the part of the Comrmssion that extensive informaten requests be carefully scrutermed by staff E

management prior to irubatino & requests. The Commissen recogrues that there may be -

instances where it is not clew whether a backfit will follow an informahon request. Those cases should be resolved in favor of analysis" GO Fed. Reg.38,112, Sep.20,1985). We beheve this I

and the language of the rule itself indicaw he Commission's original intent that Sechon 50.54(f) be used only for the most segruficant issues, when the Commission must determme whether or not the license of a facehty "should be modified, suspended, or revoked" (10CFR50.54(f)).

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Response 10CFR50.54(f) states: "....Except for informaton sought to verify licensee compliance with the current licensing basis for that facility, the NRC must prepare the romson or reasons for each informaton request prior to issuance to ensure that the burdea to be imposed on j

rest i xlents is justified in view of the potential safety significance of ti e issue to be addressed in j

the requested information......"

1 In view of the fact that the information is needed to verify licensee compliance with at I

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5 licensing basis for their facilities and that the issue is safety significant, the staff is not required to address, prior to issuance of the GL, the question of imposing burden on respondents to fumish the requeste, information to the NRC.

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'l CRGR REVIEW PACKAGE PROPOSED ACTION:

Is' sue a generic letter to request that addressees (1) assess the susceptibility of their emergency core cooling systems (ECCSs) to common cause failure as a result of reactor coolant system (RCS) draindown while in a shutdown condition, and (2) submit certain information, pursuant to Section 50.54(f) of l

Title 10 of the Code of FederalRegular/ons (10 CFR 50.54(f)),

concerning their findings regarding potential pathways for inadvertent RCS drain down and the suitability of configuration control, operating practices and maintenance procedures during reactor shutdown cooling. This information will enable NRC staff to verify whether addressees comply with NRC regulatory requirements and conform with current licensing bases for their facilities, including the establishment of, and conduct of activities affecting quality according to, documented procedures, per Criterion V of Appendix B to 10 CFR Part 50. The staffis specifically concerned about the quality control of activities (for example, the methods used to verify valve position, the controls in place to assure compliance with the plant operating procedures, and the adequacy of operator training for such activities) conducted during hot shutdown conditions affecting the safety-related functions of the residual heat removal (RHR) system and the emergency core cooling system (ECCS), as defined in 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 34 and 35, respectively.

CATEGORY:

(2)

RESPONSE TO REQUIREMENTS FOR CONTENT OF PACKAGE SUBMITTED FOR CRGR REVIEW Question m:

The proposed generic requirement or staff position as it is proposed to be sent out to licensees. Where the objective or intended result of a proposed generic requirement or staff L

position can be achieved by setting a readily quantifiable standard that has an unambiguous relationship to a readily 4

measurable quantity and is anforceable, 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.

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Response

All licensees are requested to submit the following information:

(1) Describe whether or not their ECC system includes such features as a common suction header (which can render the ECCS susceptible to common-cause failure, as a result of i

events similar to the Wolf Creek RCS drain <fown event of September 17,1994); and (2) describe, with consideration of plant specific design attributes, the features of their Appendix B quality assurance program (for example, the methods used to verify valve position, the controls in place to assure compliance with the plent operating procedures, and the adequacy of operator training for such activities) that provide assurance that the safety related functions of the RHR system W ECCS will not be adversely aNected by activities conducted at hot shutdown (such as occurred at Wolf Creek).

Licensees may limit their attention to those maintenance and operational activities at hot shutdown during which it is feasi-ble to divert RCS fluid to the RWST, resulting in simultaneous drain down of the RCS and voiding in the ECCS suction head-er. Licensees may further limit their attention to configu-rations and conditions that are risk-significant. Risk-signifi-cant configurations and conditions of most concern include MODE 4 operations involving potential flow paths that include pipe diameters equal to or greater than 2 inches, and with two or fewer normally closed valves.

Question fiil:

Draft staff papers or other underlying staN documents supporting the requirements or staN positions. (A copy of all materials referenced in the document shall be made available upon request to the CRGR staN. Any Committee member may request CRGR staff to obtain a copy of any reference material for his or her use.)

Response

e The staN issued information Notice (BN) 9543 to notify alllicensees of the Wolf Creek event.

The staN issued a supplement to IN 9543 to give licensees additionalinformation.

The Office for Analysis and Evaluation of e

Operational Data (AEOD) lasued AEODfS9541,

" Reactor Coolant System Blowdown at Woir Creek on September 17,1994," an evaluation of the Wolf Creek event and similar events.

Question (iii):

Each proposed requirement or staN position shall contain the sponsoring office's position as to whether the proposal would increase requirements or staff positions, implement existing

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3 requirements or staN positions, or would relax or reduce existing requirements or staN positions.

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Response

This is a request for information which will enable NRC staff to verify whether addressees comply with NRC regulatory requirements and conform with current licensing bases for their facilities, including the establishment of, and conduct of activities affecting quality according to, documented procedures, per Criterion V of Appendix B to 10 CFR Part 50.

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The staff is specifically concerned about the quality control of i

activities (for example, the methods used to verify valve position, the controls in place to assure compliance with the plant operating procedures, and the adequacy of operator i

training for such activities) conducted during hot shutdown conditions affecting the safety-related functions of the residual heat removal (RHR) system and the emergency core cooling 4

system (ECCS), as defined in 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 34 and 35, respectively. There are no additional or reduced staff requirements.

i Question fiv):

The proposed method of implementation with the concurrence (and any comments) of OGC [ Office of the General Counsel] on the method proposed. The concurrence of affected program j

offices or an explanation of any nonconcurrence.

Response

I This generic letter contains only information collection requirements. This information will enable NRC staff to verify whether addressees comply with NRC regulatory requirements and conform with current licensing bases for their facilities, including the establishment of, and conduct of activities affecting quality according to, documented procedures, per Criterion V of Appendix B to 10 CFR Part 50. OGC has no legal objection to this generic letter.

Question (vi:

Regulatory analyses conforming to the directives and guidance of NUREGIBR4058 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.S.(ix)).

Response

This item is not applicable to this generic letter.

Question (vi):

Identification of the category of reactor plants to which the generic requirements or staff position is to apply (that is, whether it is to apply to new plants only, new OLs only, OLs t

after a certain date, OLs before a certain date, all OLs, all plants under construction, all plants, all water reactors, all 1

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4 PWRs only, some vendor types, some vintage types such as BWR 8 and 4, jet pump and nonjet pump plants, etc.).

Response

The proposed generic letter is applicable to all PWR nuclear power plants (except those that have been amended to a possession only status).

Question fvill:

For backfits other than compliance or adequate protection backfits, a backfit analysis as defined in 10 CFR 50.109. The backfit analysis shall include, for each category of reactor plants, an evaluation which demonstrates how the action should be prioritized and scheduled in light of other ongoing regulatory activities. The backfit analysis shall document for consideration information available concerning any of the following factors as may be appropriate and any other information relevant and material to the proposed action:

(a) Statement of the specific objectives that the proposed action is designed to achieve; (b) General description of the activity that would be 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; (d) Potentialimpact on radiological exposure of facility employees and other onsite workers; (e) installation and continuing costs associated with the action, including the cost of facility downtime or the cost of construction delay; (f) The potential safety impact of changes in plant or I

operational complexity, including the relationship of proposed and existing regulatory requirements and staff positions; (g) The estimated resource burden on the NRC associated with the proposed action and the availability of resources; (h) The potentialimpact of differences in facility type, design, or age on the relevancy and practicality of the proposed action; (I) Whether the proposed action is interim or final, and if interim, the justification for imposing the proposed action on an interim basis;

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(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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1. The proposed priority or schedule,

2. A s.ummary of the current backlog of existing requirements

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' awaiting implementation, I

3. An assessment of whetherimplemoetion of existing requirements shoe'1 be deferred as a result, and

4. Any other in'armation that may be considered appropriate with regard to priority, schedule, or cumulative impact. For example, could implementation be delayed pending public comment?

Response

This is a 10 CFR 51 4(f) request for information to ensure licensee compliant s and is not a backfit; thus, a backfit i

analysis is not required.

Question fvlii):

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 (i) and (vii) above, that:

(a) There is a substantialincrease in the overall protection of public health and safety or the common defense and i

security to be derived from the proposal; and (b) The direct and indirect costs of implementation, for the facilities affected, are justified in view of this increased protection.

Response

This item is not applicable to this generic letter.

- Question fix):

For adequate protection or compliance backots 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.

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

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in Ill.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

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alternatives.

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Response

This item is not applicable to this generic letter.

Question (x):

For each ovaluation conducted for proposed relaxations or decreast.s in current requirements or staff positions, the proposi sg Office Director's determination, together with the rationc'.e for the determination based on the considerations or paragraphs (i) 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, i

and (b) The cost savings attributed to the action would be substantial enough to justify taking the action.

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Response

This item is not applicable to this generic letter.

Question fxi):

For each request for information under 10 CFR 50.54(f) (which is not sub}ect to exception as discussed in Ill.A) an evaluation j

that includes at least the following elements:

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(a) A problem statement that describes the need for the i

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.

Response

This is a 10 CFR 50.54(f) request for information.

(a) The Wolf Creek draindown event represents a PWR vulnerability that had not been evaluated previously. Staff and licensee analyses have shown that ECCS accident mitigation capability may be lost because of such a draindown. AEOD characterited the event as the most significant precursor event of 1994. The conditional core damage probability estimated in

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h NUREG/CR 4674, Volume 21, Appendix D, is 3.0E-3. In addition, the draindown creates a pathway for containment j

bypass. The information requested will help the staNidentify

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plants in need of remedial actions to avoid similar draindown events.

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(b) To gather and assess hardware and procedural j

information, a licensee may require 10 sta# days. However, licensees who have modified their systems ur their 3

procedures, may need less time for their response.

i (c) The present action plan anticipates that this generic letter will be issued one month after CRGR approval. The sta#

projects that evaluation of the responses to this generic letter

- will be completed within three months of the required submittat date. Further required inspecGons, if any, will then be completed within the following six months. At this time, the staN anticipates that the date for completing this plan will be December 31,1998.

(d) The present request is a generic letter based on 10 CFR 50.54(f) that does not impose any new requirements on the licensees other than gathering of the information.

Question fxii):

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

Ressoase:

The intent of the proposed generic letter is to assess licensees

• compliance with their facilities current licensing bases and, as such, is consistent with the Commission's Safety Goal Policy Statement.