ML20248B355

Informs Commission of Staff Intent to Issue GL 98-XX, Loss of Rc Inventory & Associated Potential for Loss of Emergency Mitigation Functions While in Shutdown Condition

ML20248B355
Person / Time
Issue date:	04/30/1998
From:	Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
SECY-98-093, SECY-98-093-01, SECY-98-093-R, SECY-98-93, SECY-98-93-1, SECY-98-93-R, NUDOCS 9806010266
Download: ML20248B355 (34)

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SUBJECT:

PROPOSED NRC GENERIC LETTER 98-xx, LOSS OF REACTOR COdfANT l INVENTORY AND ASSOCIATED POTENTIAL FOR LOSS OF EMERGENCY MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION l

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-water reactors to make available to the NRC certain information (pursuant to Section 50.54(f) of Title 10 of the Code of Federa/ Regulations (10 CFR 50.54(f))] regarding the subject matter of this generic letter. 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, 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:

The staff 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 incidc.? at the Wolf Creek nuclear power plant involving 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 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 re. cult in a loss of emergency core cooling system capability, and also involves the potential for containment bypass. Another important aspect of this event is the short time available to the operators for taking corrective action.

CONTACT: M. Razzaque, SRXB/DSSA 415-2882 TO BE MADE PUBLICLY AVAILABLE AFTER GL IS ISSUED 9006010266 980430 PDR SECY PDR .98-093 R

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The Commissioners 2 The staff proposes to issue this generic letter to request that addressees determine if their emergency core cooling systems (ECCSs) are susceptible to common-cause failure as a result 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 of such events be made available to NRC. This generic letter was endorsed by the Committee to Review Generic Requirements (CRGR) during its meeting (Number 291) on,SS2t.gmber 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 Janua'r y 22,1997, directed the staff to allow the public at least 30 days to comment before the generic letter was issued. As a result, a notice of opportunity for public comment was published in the Federa/ Registeron February 14, 1997. In response to the substantial public comments recerved, the staff revised the generic 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 susceptible to common-cause failure as a

- result of events similar to the Wolf Creek reactor coolant system draindown event; and if this  !

susceptibility is found, (2) prepare, with consideration of plant-specific design attributes, a description of the features of their Appendix B quality assurance program that provide assurance that the safety-related functions of the residual heat removal system and ECCS will not be adversely affected by activities conducted at hot shutdown (such as occurred at Wolf Creek). If the assessment performed in response to part (1) of the above requested information reveals that the susceptibility exists, then the result of the assessment must be submitted to NRC. The l response to part (2) of the above information request need not be submitted to NRC, but must i be kept in a retrievable licensee system that NRC can verify on an as-needed or sample basis.

l The staff will prepare guidance for the inspectors who will perform these verifications within the l currently available resources.

The CRGR reviewed this revised generic letter during its meeting (Number 314) on January 30, 1998, and the staff has incorporated the comments made by the CRGR at that meeting. The

' CRGR has endorsed the proposed final generic letter without formal review. The Advisory Committee on Reactor Safeguards (ACRS) reviewed this revised generic letter during its 446th meeting on November 6,1997. An ACRS letter report, dated November 13,1997, l recommended that the proposed final generic letter be promptly issued. The Office of the General Counsel has reviewed this generic letter and has no legal objections to its content.

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

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. oseph Callan 41 g l cutive Directo i

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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3 DISTRIBUTION:

Commissioners OGC OCAA OIS ---

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ACRS CIO CrO EDO REGIONS SECY l

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UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. . 20555-0001 2- May xx,1998 ,a_

NRC GENERIC LETTER 98-xx: LOSS OF REACTOR COOLANT INVENTORY ANDI ASSOCl-ATED POTENTIAL FOR LOSS OF EMERGENCY '

MITIGATION FUNCTIONS WHILE IN A SHUTDOWN CONDITION Addressees  :

All holders of operating licenses for pressurized-water reactors (PWRs), except those who have permanently ceased operations, and have certified that fuel has been permanently removed from the reactor vessel.

Purpose 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 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 certain information, pursuant to Section 50.54(f) of Title 10 of the Code of Federa/ Regulations (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 i configuration control during reactor shutdown cooling. The requested information will enable NRC staff to verify whether addressees comply with NRC regulatory requirements and conform 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' conirols over the conduct of activities during hot shutdown conditions that may affect the safety-related functions of the RHR system and the ECCS, 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 l the adequacy of operator training for such activities.

Discussion l

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, operators were attempting to reborate 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 reborated by recirculating water through a loop that contains the RHR system piping, the refueling water storage tank (RWST), a containment spray pump, a manual RWST isolation valve, and an RHR

.~ l GL 98-xx  ;

May 19,1998 Page 2 of 5 system crossover line. When the RWST isolation valve was opened for the reboration process and the train A-to-train B c rmover isolation valve was opened'for stroke testing, a drain-down path was inadvertently cres.ed from the RCS to the RWST.

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At WIlf Creek, all RHR and ECC system pump suction lines are tied into a common s header. When the draindown event occurred, hot RCS wafer wds 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 j

. operators had not been able to terminate the event, the hot water in the RWST suction piping i might have led to steam binding, which could have adversely affected the pumps in both ECCS 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.

Finally, steaming through the RWST establishes a containment bypass path. l The licensee estimated (using actual plant conditions) that for an unmitigated event, the reactor  !

l vessel water level could have drained to the bottom of the hot leg within 5 minutes and, as a j consequence, RHR pump 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 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 />.

I Events of this nature are considered particularly significant because they can result in loss of 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 l . NRC has also issued a number of other communications describing events at reactor facilities i involving inadvertent loss of reactor coolant inventory while the reactor was in a shutdown 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.

l 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 91-42, " Plant Outage Events involving Poor Coordination Between Operations and Maintenance Personnel During Valve l Testing and Manipulations," dated June 27,1991, the NRC discussed inadvertent loss-of-l inventory events. The AEOD report " Reactor Coolant System Blowdown at Wolf Creek on l 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 coordination between operations and maintenance personnel, and inadequate procedures associated 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 procedural deficiencies were related to omissions or lack of specificity in sequential valve operations when conducting tests on the RHR system. On the basis of this history and the potential 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 function capability of the RHR and ECC systems.

w-_________=_--____-_______--___-____ _ -- --__ -__ _ ._ .

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GL 98-xx May 19,1998 Page 3 of 5 Reauired Information Within 180 days of the date of this generic letter, addressees are required to perform the following (1) an assessment of whether your emergency core cooling systems jgelyde_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 dimilar to the Wolf Creek R'CS drain-down event of September 47,1994; 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 functions of the RHR system and ECCS will not be adversely affected by

. activities 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.

Addressees 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 part (1) of the above required information reveals that the susceptibility exists, then the result of the assessment shall be submitted in writing, {

pursuant to 10 CFR 50.54(f) and 10 CFR 50.4, to the U.S. Nuclear Regulatory Commission, 1 ATTN: 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 in a retrievable licensee system that NRC can verify on an as-needed or sample basis.

Baekfit Discussion L This generic letter only requests information from the addressees under.the provisions of l

Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f), to verify addressee compliance with the Commission's regulations and conformance with the current j

3 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 B to 10 CFR Part 50, the requested information will enable the NRC staff to i determine whether adequate control is being exercised over surveillance, maintenance,  !

modification and operational activities conducted at hot shutdown which can adversely affect

. the safety-related functions of the RHR and ECC systems. No backfit is either intended or approved in the context ofissuance of this generic letter. Therefore, the staff has not performed a backfit analysis.

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GL 98-xx May 19,1998 Page 4 of 5 Quality Assurance Criterion V of Appendix B to 10 CFR Part 50 requires that " activities affecting quality shall be prescribed by documented instructions, procedures, or dravings of a type l appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings." Furthermore, licensees' technical specifigggns_ include  !

requirements to establish, implement, and maintain written administrative procedures to address startup, operation, and shutdown of a shutdown cdoling system. Maintenance and testing activities at Wolf Creek during hot shutdown were carried out contrary to documented procedures and the technical specifications, resulting in RCS drain-down and the potential for ,

common-cause failure of the RHR and LCC system pumps, which could have compromised the  !

ability of the RHR and ECC systems to fulfill their safety functions. Furthermore, the staff has determined that similar loss-of-coolant events while on RHR cooling have occurred at over 19 l

plants. These events were due to the failure on the part of licensees to either establish I 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 l 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 i capability of the RHR and ECC systems. In accordance 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 Notification A notice of opportunity for public comment was published in the FederalRegister(62 FR 7075) on February 14,1997. Comments were received from four nuclear utility companies, the Nuclear Energy institute, and the Nuclear Utility Backfitting and Refo,m Group. The staff's evaluation of the comments is available from the NRC Public Document Room. The generic letter has been appropriately revised to reflect the comments received.

Panetwork Reduction Act Statement This Generic Letter contains information collections that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved i

by the Office of Management and Budget, approval number 3150-0011, which expires -

September 30,2000.

l The public reporting burden for this mandatory information collection is estimated to average L 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 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 potential impact of the information collections contained in the generic letter and on the following issues:

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GL 98-xx May 19,'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 pragt utility?

2. Is the estimate of burden accurate? * --

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

L 4. 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 L 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.

l 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@nre. gov Lead Project Manager: Kristine Thomas, NRR (301) 415-1362 L

E-mail: kmt@nrc. gov

Attachment:

List of Recently issued NRC Generic Letters

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i Attachment 2 PUBLIC COMMENT RESOLUTION AND STAFF RESPONSE Substantial comments were received from public/ industry sources as a result of publishing the proposed Wolf Creek generic letter (GL) in the Federal Register. These comments have been consolidated by the staff, and are presented below. The original letters from the pti6ffc/ industry with their comments are attached herewith.

Comment # 1: [SCE&G/W&S/NEl/TVA/CP&L/BGE] The proposed GL ignores licensee reviews and any action taken regarding previous NRC notifications (IN 95-03 & Supplement, NUMARC L 91-06: " Industry Guidelines to Assess Shutdown Management," and Generic issue 105:

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" interfacing System LOCA in LWR"). Corrective actions have been taken in the form of estab-lishing appropriate work controls to preclude valve misoperations during shutdown conditions.

We recommend the NRC Resident inspectors (RI) appraise the operating experience evaluation of the !N and other generic communications, and any utility review / action taken regarding the Generic issue, instead of issuing a GL. Data collected by this inspection could be added to the next RI's monthly report. These actions would quickly provide the desired information 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. Altematively, the NRC should provide licensees at least 180 days or until the next refueling outage to respond to the GL.

Some (but not all) PWR plant designs utilize a common ECCS suction header and are potentially susceptible to this type of failure mode given valve misoperations during shutdown. PWR plants are familiar with the sequence of events described in the IN and the AEOD Special Report on the Wolf Creek event. BGE claims that it's plant design does not allow one shutdown cooling loop aligned to the RCS with a second loop aligned to the RWST, and that the design also includes separate suction headers from the RWST to the ECCS pumps. As a result, these plants provide an adequate basis for assuring that the ECCS is not subject to common-cause failure in the I event of.an RCS drain-down while in a shutdown condition. BGE, therefore, claims that the plants having such design should not be subject to the proposed GL. They recommended that NRC further research the appropriate applicability of this concem prior to issuing a GL applicable to all nuclear power plants.

4 The action requested by the proposed GL represents a major burden on licensees that is neither practical nor necessary. Plant procedures are in place to effect administrative controls addressing normal shutdown and testing valve alignments. Categorization and description of the i

permutations of valve lineups and potential misoperations that could contribute to RCS draindown events would entail an exhaustive effort with little practical significance. It is certainly '

important that each outage operation involving these valves be carefully reviewed with regard to l draindown potential at the time that the outage work is planned or executed. However, to attempt 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 i

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. j 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 .

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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2 Response: If the licensees whose plants are susceptible to Wolf Creek like event, have indeed 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 available in 120 days. Any corrective actions taken are required to be documented and reported to appropriate levels of management in accordance with Section XVI of Appendix B to 10 CFR Part 50. This kind of request for informatic; by NRC is not done routinely, but only in special circumstances such as the Wolf Creek event, which was the most significant p, Mor event of '

1994 with a conditional core damage probability estimated to be 3.0E-3. However, in order to further relax the requirements, the GL has been revised such that it requests information at this time, and no backfitting will be required. Furthermore, the time allowed to prepare the responses has been increased to 180 days. In the revised GL, the staff requests that addressees 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 functions of the RHR system and ECCS will not be adversely affected by activities conducted at hot shutdown (such as occurred at Wolf Creek).

If the assessment performed in response to part (1) of the above required information reveals that the susceptibility exists, then the result of the assessment must be submitted to NRC And, if the assessment does not reveal that a susceptibility exists, then no submittalis necessary.

However, responses to parts (1) and (2) of the required information shall be kept in a retrievable licensee system that NRC can verify on an as-needed or sample basis.

The staff agrees that the scope of the requested information in the proposed GL was too broad, and that a more focused review would be sufficient to address the issue. Based on the Wolf Creek experience, the staff believes that the configurations, conditions and processes during shutdown which are most risk-significant must be addressed. Hence, the " Required information" section of the GL has been modified to read: " Addressees may !imit 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 voidiag in the suction header for the RHR and ECC system pumps. Addressees may further limit their response to the consideration of potential configurations and conditions that involve flow patfis with pipe diameters equal to or greater than 2 inches."

Comment # 2: [NE!/W&S/TVA/CP&UBGE] Plants have already demonstrated and been licensed on the basis of compliance with GDCs 34 and 35, and design features alone cannot preclude this or similar 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 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 relative to outage conditions and should be subject to regulatory analysis.

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

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3 inappropriately characterized as compliance exceptions to the backfitting provisions of Section 3 4 50.109. The requirements of Section 50.46 relate to specific design features of the ECCS, whereas the problem described in the proposed GL relates to "the adequacy of ECCS configu-ration control and operating practices." The actions requested by the proposed GL more appropriately relate io 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 10CFRSO.109, the proposed GL implies that the ECCS must be designed to preveMuch scenarios, when licensees generally rely on administrative controls to p,revent placing the RCS and ECCS in such a configuration.

According to the NRC's SRP, the ECCS is designed to refill "the vessel in a timely manner for a LOCA resulting from a spectrum of postulated piping breaks within the reactor coolant pressure boundary"(Sec.15.6.5, NUREG-0803, Rev.2, July '81). Even though operator actions may result in a potential pathway for loss of reactor coolant inventory during shutdown conditions, the consequences are not commensurate with a pipe break at full power operations and modifica-tions to the design features of the ECCS may not be the most appropriate corrective actions to address this situation.

The NRC Staff position implied in the proposed GL 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.

Response: The Wolf Creek-like scenario is credible for some PWR plants that utilize a common ECCS suction header. The facility may have been designed in accordance with GDC 34 and 35, as the comments claim, but if Wolf Creek-like event occurs, there exists the potential that the functions as defined in the technical specifications (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 activities (for example, the methods used to verify valve position, the controls in place to assure compliance with the plant operating proce-dures, and the adequacy of operator training for such activities) conducted during hot shutdown ,

conditions affecting the safety-related functions of the RHR system and the ECCS, as defined in 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 34 and 35, respectively. Criterion V of Appendix B to 10 CFR Part 50 requires that " activities affecting quality shall be prescribed by docuru c ted instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings."

Furthermore, licensees' TS include requirements to establish, implement and maintain written administrative procedures to address startup, operation and shutdown of a shutdown cooling sysfe~m. Maintenance and testing activities at Wolf Creek during hot shutdown resulted in the RCS drain-down and the potential for common-cause failure of the ECCS pumps, which could have compromised the ability of the RHR and ECCS systems to fulfill the safety functions specified in GDC 34 and 35, respectively.

The GL has been revised to request only information from the addressees under the provisions of 10 CFR 50.54(f). In view of the Wolf Creek draindown event, this information is needed to verify licensees' compliance with NRC regulatory requirements and current licensing bases for their facilities as related to the requirements of Criterion V of Appendix B to 10 CFR Part 50, specifically as regards the quality control of activities which can adversely affect the safety-related functions of the RHR and ECC systems, whose requirements are defined in 10 CFR Part 50, Appendix A, GDC 34 and 35. i Comment # 3: [CP&L) Although it appears from the referenced events that the generic letter is

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4 specifically concemed 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 failure of PWR ECCSs or susceptibility of l PWRs to drain-down events.

Response: The GL has been clarified in the " Required Information" sectica which states,

" Addressees may limit their attention to those surveillance, maintenance, modificatiMand opera-tional 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. Addressees 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."

Comment # 4: [W&S) Management controls for work activities in shutdown operations, when property 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 importance of proper administrative controls by the generic communications 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 information and any new interpretation of the regulations. If the staff believes that it has addi-tional information or insights useful to licensees, a second supplement to IN 95-03 could be issued rather than the proposed GL Response: The GL requirements have been relaxed to request information only, pursuant to 10 CFR 50.64(f), and that backfitting is not required by the GL at this time. However, NRC needs the information, as discussed in the GL, to 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.

Comment # 5: [W&S) in the Statement of Considerations for the revision of Section 50.54(f), the NRC states that "if extensive effort is reasonably anticipated, the request will be evaluated to determine whether the burden imposed by the information request is justified in view of the potential safety significance of the issue to be addressed...... Requests for information to determine compliance with existing facility requirements...... usually are not made pursuant to 50.54(f)...... The amendment of 50.54(f) should be read as indicating a strt6g concem on the part of the Commission that extensive information requests be carefully scrutinized by staff mafagement prior to initiating such requests. The Commission recognizes that there may be s instances where it is not clear whether a backfit will follow an information request. Those cases should be resolved in favor of analysis" (50 Fed. Reg.38,112, Sep.20,1985). We believe this and the language of the rule itself indicate the Commission's originalintent that Section 50.54(f) be used only for the most significa.nt issues when the Commission must determine whether or not the license of a facility "should oe modified, suspended, or revoked" (10CFR50.54(f)).

Response: 10CFR50.54(f) states: "....Except for information sought to verify licensee compliance with the current licensing basis for that facility, the NRC must prepare the reason or reasons for each information request prior to issuance to ensure that the burden to be imposed on respondents is justified in view of the potential safety significance of the issue to be addressed in the requested information......"

In view of the fact that the information is needed to verify licensee compliance with the current

r. .

~

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 requested information to the NRC.

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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 Loss of Emergency Mitigation Functions While in a Shutdown Condition" (62 Fed. Reg. 7075 - February 14,1997)

_Oonortunity for Public Conunents The Nuclear Energy Institutet offers the following comments relative to a Federal Register notice which solicited public comments on a proposed NRC i

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.

i The proposed GL addresses concerns resulting from the Wolf Creek RCS drain-down event of September 17,1994 (NRC Information Notices 95-03 and 95-03, Supplement 1). Licensees are requested 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 pursu' ant 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

( mergency core cooling) of Appendix A to 10 CFR Part 50.

t I

NEl is the organization responsible for establishing unified nuclear industry policy on matters affecting energy industry, including regulatory aspects of generic operational and technical issues. NE! members include a utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant design architect / engineering firms, fuel fabrication facilities, materials licensees, and other organizations and individu involved in the nuclev energy industry.

[

U. S. Nuclear Regulatory, Commission

-March 17,1997 ~

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 / )

operation of the ECCS and residual heat removal system (RHRS). _In oractice.

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

[ During cold shutdown and refueling conditions, the Technical Specification

/ y 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 permutations of valve lineups that /

/

p  ! 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. na ition, the outage plan is con 6gured and controlled to ass";e the performance of key safety functions. Plants are committed through 10 CFR 50 Appendix B and their Technical Specifications to establish and follow procedures 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. Q6 We believe this has been accomplished through issuance of

" ndustry Guideline to Assess Shutdown Mana em t " an industry reviews C 91-06 y/

i of eI o ation otices res mg from t e Wolf Creek event. Further, we believe the action requests contained in the proposed GL are unnecessary given the work controls discussed above. l y (I'he proposed GL discusses a potential common-cause failure mode involving 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

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recirculated Duid to enter the shutdown cooling system between the borated

, water storage tank outlet and pump suction. These plants are not susceptible 8% ,

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 G

" susceptibility" to this failure mode. Susceypility,to,this. common-cause failure

. mode is predicated. upon errors of commission (vnive 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  !

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 siministrative controls during shutdown. This has been accomMd. i l througl implementation of the industry initiative associated with

' NUMARC 91-06 and further emphasized through industry awhreness 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 WolfCreek RCS, drain- n event.

Comment: Some (but not allyPWR plant designs utilize a com on E CS suction b-9: md - ntiall s ce i e s o all e mode -

give w2 ve o

. r-ananca ar :.

s su wn. WR plants are familiar describe in the Information Notices and the s[  !

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 .

accordance with the requirements ofSection XVI ofAppendix B to 10 CFR

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

Comment: Plants have already demonstrated and been licensed on the f basis of compliance to GDCs 34 and 35, and design features alone cannot

. preclude this or simil s'.-C6ifesti7ve a' ns have been taken in the (gj, form of establishi appropriate work controlslo preclude valve misoperations duri . - d ~cc . ce a m M '

Action request: If the RCS is found to be susceptible to drain down events, describe each puential 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, and describe any administrative controls that are intended to be used to control valve manipulations to preclude RCS drain-down events.

___ _ _ _ _ _ . _ _ _ ____ ____.______-___.___m____________________-___________________.___._____.m_ _ . _ . _ _ _ _ _ . . _ . _ . _ _ _ . _ _ _ . _ _ _ . _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ _ _ _ _ _

U. S. Nuclear Regulatory Commission March 17,1997 i

Page 4 . I t

Comment: This action request represents a major burden on licensees that is neither practical nor necessary. Plant procedurcs are in place to, effect administrative controls addressing normal shutdown and testing ~

valve abg,qats. Categorizationliid~descEjition of thETerin'utations of h

valve lineups and potential misoperations that could contribute,tafSS_ '

" drain-down events would entail an exhaustive effort with little practical significance. It is certainly important that eacil outage dperation 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_

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

With regard to the backfit discussion of the proposed GL. we offer the following- [

j If the intent is to interpret 10 CFR 50.46 and the relevant GDCs to suggest that f f design features must be in place to preclude these types of events during shutdown operations, then we believe the proposed GL represents a significant 1

. l' backfit requiring a regulatory analysis. Further, if the intent is for licensees to -

investigate and categorize all permutations of valve operations that could lead 3'. .';

to this type of event, we believe this represents a significant revision to current t practices for regulation relative to outage conditions and should be subject to j 1 regulatory analysis.

We appreciate the opportunity to provide comments on this proposed Generic Letter. Ifyou have any questionsi regarding these comments, please contact Biff Bradley at (202)-739-8083.

Sincerely, .

Y -

OYQ Anthony R. Pietrangelo REB /ec

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CHARLES H. CRL SE ., ..

8J!!imOre CsJs and Electne Company y;ce p4ident Cahen Chtf Nuclear Power Plant Nuclear Enery)

M50 Caheri Chtfs Parku a3

l Lusby. Margand 20657 y Gr '

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l March 18,1997 f

i U. S. Nuclear Regratory Commission Washington, DC 20555 i

(

ATTENTION: Rules Review and Directives Branch i

SUBJECT:

Calvert Cliffs Nuclear Power Plant' I 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" 4 1

I 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 eneric letter would: (1) assess the susceptibility of their Emergency Core Cooling System to common-cause failure as a result of Reactor Coolant System drain-down while in a shutdown condition; and (2) submit certain information concerning their findings regarding potential pathways for inadvertent Reactor Coolant System drain-down, and the suitability of configuration control and operating practices during rerf, tor shutdown cooling.

We feel that the proposed generic letter should not be issued. We feel BGE's existing design provides an i 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 j

Calvert Cliffs. Baltimore Gas and Electric Company's design does not allow one shutdown cooling loop l aligned to the Reactor Coolant system, with a second loop aligned to the refueling 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. Calvert Cliffs and similarly designed plants '

should not be subject to the proposed generic letter. _We recom_ mend the Nuclear Regulatory Commission further research the appropriate~ applicability

~~^

of this concern prior to issuing a generic letter applicable to all nuclear power plants. ~~ ~ ~ ~ " " ~ ~ ~ " ~ ~ ~ " . ,

1 Rules Review and Directives B' ranch March 18,1997 Page 2 ,

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

Very truly yours,

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cc: Document Control Desk, NRC H. J. Miller, NRC D. A. Brune, Esquire Resident Inspector, NRC J. E. Silberg, Esquire R. I. McLean, DNR Director, Project Directorate I-1, NRC J. H. Walter, PSC A. W. Dramerick, NRC l

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Washington, D.C. 20555 0001 0

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RE:

Proposed Generb Letter, "Less of Rasetor Coolaat Inventory and Associated Potential for Loss of e-wf Mitigstlos Fumations While la a shutdown

.C,gg did ag ,H J 2 W ed. n 7,393 g & _ _.,14, iggan ATTN: Chief, Rules Review and Dimetives Banch on February 14,1997, the Nuclear Rasulatory waaina ("NRC") lasued the d aW proposed Oensric Lemur forpublic maat Provided below arethe comunents F 'these aama=wa concern the 1

In the proposed Gunsde I. meter, the NRC requests licaa==== *) (1) assess the susceptibility of their emergency oore cooling systems ("ECCS") to comanon cause failure as s l

result ofmastor coolset system ("RC3") drain down while in a shualown condition, and (2) sub l

lae==adaa pursuant to 10 C.FA 950.54(f) concoming their W sogarding poesntial pathwa forinadvertent RCS dr.da-down ".d the udtability of onguration control ami openting pracdcas

< during sessent shumiown Wian i The la#maanian is to enable the NRC Staff to verify wbsther addressoas comply and conibar, with NRC regulatory and license aquimmensa, specifloally verilpas the adequacy of mainamining the residual heet removal asisty fbaction to transfer produst doesy heat and other residual heat from the roastor (Geomal Design cdterion M of Appendix A to Part 50) and the ECC5 to provide abundant emergency core cooling when requi (General Design Criession 35).

\

.DissJerticular event described inthe pregosed Genede Letter eannaa=1 - - - . d, af a mactivities rather man .,

perfbrmed r=- - aan sc+wf. Certain eactfWii~ s ws s ongoing Gi, resulted in an open drain < lown path from the RCS to the rufustfag water is a, consortium of 15 utilities formed in the early 1980s which participated

@is act in the development of the NRC's backfitting rule (10 C.FA 450.109)in 1985, and which.....

has-. closely monitor,ed the NRC's application of the rule since that time.

[_ _

G. thib .gki ~ " '- -- - -

WINSTON ds STRAWN ^*

i U. S. Nuclear Regulsory Commission March 17,1997 Page 2 storage tank. "Ihe "Backfit Discussion" of the proposed Generic Letter refers to compliance with 10 C.F.R. l50.46, " Acceptance Criteria for Erw ;y Core Cwitag Sy *- =fsrlight Water Nuclear Power Pe." to justify the iafa==% :equest to eaneren the adequacy of existing /

ECCS con 8guration control and operating Ma* ra**=lia! ssidua'l heat removal. The "Backfit Discussion" of the proposed Generic Letter stsess:

The actions auquested in this generic letter, if required, would be backfits in accostlance with NRC procedutos and are necessary to enano that addromoes are in compliance with existing NRC rules and regulations. Sp-shny,10 CFR 50.46 requhes that the ECCS be desisned to y... ids adequaes Sow capability to maintain the core tempersons at an acceptably low value and to tornovo (lacey hast for the extended period of time r==sai by the long. lived radica.,;M;y remaining in the core. 'the Wolf Creek event has damanan*=t that the adequacy of ECCS

=adenstion contal and operating ;-ems sosseding residual hast removal can adversely impact ECCS per rn,---. and could prevent the ECCg Bom personning its asisty fbaction fbliowing events at reactor facilities involving inadvertent loss of reactor coolant . ; i while the reactor is shut down. Therefore, this generic letter is being issued as if the requested actions were compliance backfits under the tanns of10 CFR50.109(aX4Xi).

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characterized as --ha- ---+r . to the Wi*i-- nrovialaa= of E- *% 50.109. ne

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requirements of Section 50.46 solase to specine design issames of the BCCS, whereas the problem daaedhad inthe psoposed Gensric Ietter salsass to "the adequacy ofECCS configuratica control and operating practices." The actions requested by the proposed Genesis Latter more appropriately relata to conduct and coordination of andvities while in a shut down candition. By citing Section /

50.46 as the bois fbr tir .omplianos exception to the backftt provisions of 10 C.F.R. 550.109, the proposed Generic IAtter kuplies that the ECCS must be designed to prevent anch scenarios, when licensees generally miy on adniinistrutive controls to prevent placing the RCS and ECCS in such a confignadon. Aeoording to the NRC's Standard Review Plan, the BCCS is designed to refill "the vessel in a thnely manner for a loss of-ooolant accident vesulting Besa a spc3xn of postulated t y l piping taeaks within the zusator coolant pressure boundary.'e' Even though opemaer actions may, '

resuh in s_3essatialpatitay for loss of reactor coolant inventairy dadng shut down ==Ma I iciglingusages_attaot aa===aiirriiisiwlETiiipe beslist fb11 power edonsoperation ,

I to the desianisstmos ofins mu s menot be_th_e.most appsprisse w sve estions to asaress tYis i , j situation. The NRC Staff position implied in the proposed Generic Letter appears to be a new i ) . l ist ,.. tion of ths -Maa= in Section 50.46 which would be subject to the backfitting i i provisions ofSection 50.109.

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P.23 WINSTON de STRAWN ~

U. 5.Nuc'carRegulacry Commission March 17,1997 Page3 Licensees have been made aware of the event discussed in the juo Letter by informadon Naties 95 03 and oth.e generio conspMaae refersnoo Osmetic Letter, f * =+-f actions taken as a result of these parvious #pdproposed !f

-- :s to nuevent inadverr.st opensor acLes IEE crema a GiL.,:sW. pahway may be laaaaa==d W  ! NRC. J1 iasaraaa maussis eneoussruy tea-a== in no v ar oeur s=%=3 = the necessary =pe sadWanyaquim w='eto The, tes tanknian ofstudies, Imr wenid ask ueensses,itthe acs is eune to n snm.pshi. to drain down =nis. mUcnene d cribe L %

• eech potsetial drain < lown flow path (include piping sians. Identi$ flow path valves and their naanat po.hians, .od idanary valve int. docks and gans ser vain ,-% indi.edan in the connoi mom), describe pomadal valve testing manipaissons or'uses, and desaihe say

% eammals that an imunded to w und to control vain manipama to precinde Res drain.down evann.a an.ii.v.e e ads ivo. .rmu.t is now-.a th, e m_ , w that es burden an &

is J- de In the sim-a-* ar ces er en avisionTr Menon 30.390, theNRC mas that "ti]f extsasive effort is nasonably anticipated, ths esquest will '

o be evaknesd a dessenine whater the butden imposed by tbs inkmmian aquesti isjestiS of the poesntial safbty significate of tbs imus to k addressed. . . . Raqosses ist intnom d*=mhm amplians with amisdag imellity requirmosas . . . umaDy su not made pumusst to,f 150.340.

e . . . The ====6a== of 150.540 should be med as indicadng a semog eenoma en en p ares -am that sammtw lawaa requma be amadir amminised by marminneensat & f prior to fattisting such toquests. The Commission necognisse that these may be instammes Iwhste it is not clear wheese a imetat wiu sonow an intonnasien agosse, nom enans should be neotv.4 in hveeofsmalysis.'*' We beliew this and the lansass ofthe rule itssifindioses the Commission's t original latent that Section 30.5(f) be und only kr the most signisaast issues when the ea==danian must datummine whether ornot the' lie.nse of a facility "should be madh_e sapended, l I

or'tevoked."8' 1 \

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implemented, pmvide a tossenshie means of reviewing possible valve combinations that could be Toadwatuntly ming-irianad during specific work activities ca a --- 'ri== basis, Lise basa made swore _oftheW ofproper ademistrative costols bythe generis h s j, i :ss==n e in enprop e o n.m. w m , ; e, er am inn. am m4 1 i Gemene Lemur dikh a == = analysis has 5=a== ===d=*=d, Justi$las the amed for ttui

~had anynewinserinstanonsofte reguistions. If ths Staff 1+'i . Getithas madman =F i

F- 50 Fed. Reg. 38,112 (September 20.1985).

8' 10 C.F.R. 550.54(f).

Ls:x::,,< tcLo- --

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WINSTON & STRAWN .~

U.S. Nuclear Regulsory Commission March 17,1997 Page 4 h or insights ussfW to licensees, a second surle to Infonnation Notice 9543 could f

~ -

be issued raherihan tfspr6pom venene Letter. _ ,<

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Sincerely, . .-

s bA A w _ _ , .

DanielF.Stunger

}

PeisisI,. Campbeu Counsel for Nasiser Utility Bukfluing and Rdnne Group

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J. Sheph (.2fg 7.975-tFAL /Z;&w/n

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Carolina Pc,wer & Light Company C PO Box 1551 411 Fayetteville Street Mall

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Raleigh NC 27602

• Serial: Py_S-97-024 March 14.1997 - -

Chief, Rules Review and Directives Branch U.S. Nuclear Regulatory Commission Mail Stop T-6D-69 Washington, DC 20555-0001 L

Subject:

Comments on Proposed NRC Generic Letter 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 Propo Generic Communication regarding " Loss of Reactor Coolant Inventory and Associated Potential for Loss of Emergency Mitigation Functions While in a Shutdown Condition." The propose letter requests that licensees " determine.whether the.r ECCSs are susceptible to common-cause failure, 17,1994."e.g., as a result ofevents similar to the WolfCreek RCS drain-down event of Septem 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, and describe a 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 ofI

/

d' the requested response is unclear . It is recommended that the generic letter be clarified as to whether it is inten hd to address common cause~ failures of PWR p-d f

ECCSs or susceptibility of PWR.s .to dr..ain.-d..own event.s.

i . The event that was the genesis of this proposed generic letter occurred in September 1994. Subsequently, the NRC issued Information Notice 95-03 and its supplement. /

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

.~

. Chief. Rules Review and Directives Branch  : .\ larch 14.149T 4

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

The proposed letter requests PWR licet. sees to evaluate the susceptibility of their I~

ECCSs to common cause failure and implement corrective actions iiiWo'rdance with 10 CFR 50. Appendix B. Criterion XVI if they are fonnd to be susceptible.' CP&L agrees that this request has merit whether based upon the NRC Information Notice or this proposed generic letter. However the second part of the letter requests licensees i to provide a large volurr.e ofinformation concerning each potential drain-down flow ,

path. CP&L questions if there is any additimal safety benefit from providing this large amount of detailed data since 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 j Murdock at (919) 546-3193.  !

Sincerely.

T.D. Walt "

. Manager, Performance Evaluation

& Regulatory Affairs MLM/

cc: Mr. J. B. Brady, USNRC Resident Inspector - HNP, Unit 1 Mr. B. B. Desai, USNRC Resident Inspector - HBRSEP. Unit 2 Mr. N. B. Le, USNRC Project Manager - HNP, Unit 1

- Ms. B. L. Mozafari. USNRC Project Manager - HBRSEP, Unit 2

$ /

I

~7 Gary J. Taylor

/ . )q #8

//Wh P.O. Box 88z South Carolina Eltetric & Gas CompIny Vice President Jenkinsvitie. SC 29065, . Nuclear Operations ~

7J2e f t'C (803) 3454344 . ..3-U

• 7#75 SCE&G ASCme:Corrpany

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March 12, 1997 y RC-97-0055 Chie4 Rules Review and Directives Branch l

_U S. Nuclear Regulatory Commission l

" Mail Stop T-6D-69 '_ ' ~

Wmchimrtan, D. C. 20555-0001 . -

Sir:

Subject VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 COMMENTS ON THE PROPOSED GENERIC LETTER ON LOSS OF REACTOR COOLANT INVENTORY AND ASSOCIATED POTENTIAL FOR LOSS OF EMERGENCY MITIGATION FUNCTIONS WHILE IN A SHUrDOWN CONDITION {

South Carolina Electric & Gas Company (SCE&G) has reviewed the proposed generic letter and has the following comments. I The proposed generic letter ignores licensee reviews 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 expe rience evaluation of the Information Notice and any utility review / action taken regarding the Generic Issue. instead ofissuinia generic letter. Data

/

'/

collected by tais mspecuon coulo De added to the next Resident Inspector's monthly report.'lhese actions would quickly provide the desired information within the time-frame of the initial 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 Genericletter.

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

Very Truly Yo W

Garyh lor cac c: J. L. Skolds W. F. Conway R. R. Mahan R. J. White L. A. Reyes A. R. Johnson

NRC ResidentInspector

! NSRC RTS (MSP 970005)

File (810.32)

Documents Management System i

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Ma_rch 14, 1997

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f Chief, Rules Review and Directives Branch U.S. Nuclea" Regulatory Commission Mail Stop Tm JD-69 .

1 Washington, D.C. 20555-0001

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Gentlemen:

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

" LOSS OF REACTOR COOLANT INVENTORY AND ASSOCIATED FOR LOSS SHUTDOWN OF EMERGENCY CONDITION" MITIGATION FUNCTIONS WHILE IN A TVA offers Reaister the following comments relative to Federal _

Notice l

l (62 FRN7075, February 14, 1997) which solicited public comments on a proposed NRC GL requesting licensees core coolingtosystems assess(ECCSs) the susceptibility of their emergency result of a reactor coolant to common system cause while drain-down failurein asaa shutdown condition.

The proposed GL requests that licensees determine whether their ECCSs are susceptible to common-cause failure as a l I

result of system events (RCS) similar to the Wo.f Creek reactor drain-down event of September 17, coolar 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 design criteria with(GDC)the regulatory 34 and 35.guidance provided in general l Comment:

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

Categorization and description of the permutations of valve l

! lineups and potential misoperations that could contribute to s/

( RCS drain-downsignificance.

little practical events would entail an exhaustive effort with I It is certainly important that each outage operation involving these valves be

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Chief, Rules Review and Directives Branch l

Page'2 March 14, 1997 carefully reviewed with regard to drain-down potential at the-time that the outage work is planned. However r '90' - l attempt to accomplish.this in advance..f.or,all_possible - /

combinations ls 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 {

onacase-by-casebasis,where-thescopecanbeconstrained{

to practical dimensions, and greater assurance provided of g

. an intensive review.

I In addition, suitable consideration of redundancy, single failures, and power supplies ensures reliable operation of the ECCSs and residual heat removat system, and Technical Specifications preclude automatic ECCSs actuation during cold shutdown conditions. Administrative and work controls must be. relied'upon for evolutions described by this event. y/

TVA utilizes a " protected train concept" as discussed in _

NUMARC 91-06, in conjunction with controls over emergent cf worn activities. The outage plan is configured and

f controlled to ensure the performance of key safety functions. -

I f We recognize the safety sig;.ificance 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 I conditions. We believe that thi. hasLessentially been

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accomplished through-implementation of NUMAnc 91-06 and y through: industry r /iews 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 Lagard~ to the backfit discussion of the proposed GL, we 1! 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

!e shutdown operations,-then we believe the proposed GL represents.a significant backfit requiring a regulatory /

analysis. Further, if the intent is for licensees to i ' investigate and categorize all permutations of valve t operations that could-lead to this type of event, we believe i this represents a significant revision to current practices for regulation relative to outage cohditions and should be

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i) subject to regulatory analysis.

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Chief, Rules Review and Directives Branch '

Page 3 March 14, 1997 i'

TVA appreciates sproposed GL. the opportunity to provide comments _on this If you comments, please contact have E.

any questions regarding-fMWEe-7 W. Whitaker. at (423) 751-6-369.

Sin rely,'

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Raul R. Baron General Manager Nuclear Assurance and Licensing cc: Mr. Ronald W. Hernan, Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852 Mr. Robert E. Martin, Senior Project Manager U.S. Nuclear Regulatory Commission 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 2900 Atlanta, Georgia 30323

_ Mr. M. C. Thadani, Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, Maryland 20852 Mr. J. F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike j

Rockville, Maryland 20852 l

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Chie.f, Rules R'eview and Directives Branch Page 4 March 14, 1997 cc: NRC Resident Inspector t -- -Browns Ferry Nuclear Plant - adLE -

10833 Shaw. Road , ,

Athens, Alabama 35611 NRC Resident Inspector' Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379 NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear-Plant Road Spring City, Tennessee 37381 N'

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, March 17.1997

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U.S. Nuclear Regulatory Commission Rules Review and Directives Branch Mail Stop T-6D-69 Washington, D.C. 20555-0001 {c u

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RE:

Proposed Generic Letter, " Loss of Reactor Coolant Inventory and Asdciath Potential for Loss of Emergency Anitiga.:sn Functions While in a Sh@down Condition." 62 Fed. Ree. 7.075 (February 14.1997) '

ATTN: Chief, Rules Review and Directives Branch On February 14,1997, the Nuclear Regulatory Commission ("NRC") issued the eve-captioned proposed Generic Letter for public comment. Provided below are the comments j

of the Nuclear Utility Backfitting and Reform Group ("NUBARG").I' These comments concern 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

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result ofreactor coolant system ("RCS") drain down while in a shutdown condition, and (2) submit information pursuant to 10 C.F.R. {50.54(f) conceming their findings regarding potential pathways for inadvenent RCS drain-down and the suitability of configuration control and operating practices ;

during reactor shutdown cooling. The information is to enable the NRC Staff to verify whether addressees comply and conform with NRC regulatory and license requirements, specifically ,

verifying the adequacy of maintaining the residual heat removal safety function to transfer fission

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product decay heat and other residual heat from the reactor (General Design Criterion 34 of

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Appendix A to Part 50) and the ECCS to provide abundant emergency core cooling when required j (General Design Criterion 35).

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I The particular event described in the proposed Generic Letter concemed management

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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 i

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1 NUBARG is a consortium of 15 utilities formed in the early 1980s which participated l

' activelyin the development of the NRC's backfitting rule (10 C.F.R. @50.109)in 1985, and which has closely monitored the NRC's application of the rule since that time.

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WINSTON & STILMVN i

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U. S. Nuclear Regulacry Commission March 17.1997 Page 2 storage tank. The "Backfit Discussion" of the proposed Generic Letter refers to compliance with 10 C.F.R. {50.46 " Acceptance Criteria for Emergency Core Cooling Systems for Light Water NucleatPower . Reactors," to justify the information request to confirm the adegacy.of existing ECCS configuration control and operating practices regarding res,idual 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 necessary to ensure that addressees are in compliance with existing NRC rules and regulations. Specifically,10 CFR 50.46 requires that the ECCS be designed to provide adequate flow capability to maintain the core temperature at an acceptably low value and to remove decay heat for the extended period of time required by the long-lived radioactivity remaining in the core. The Wolf Creek event has demonstrated that the adequacy of ECCS

. configuration control and operating practices regarding residual heat removal can adversely impact ECCS perfonnance and could prevent the ECCS from performing its safety function following events at reactor facilities involving inadvertent loss of reactor coolant inventory while the reactor is shut down. Therefore, this generic letter is being issued as if the requested actions were compliance backfits under the tenns of 10 CFR 50.109(aX4Xi).

We believe that actions requested in the proposed Generic Letter are inappropriately .

characterized as compliance exceptions to the backfitting provisions of Section 50.109. The

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requirements of Section 50.46 relate to specific design features of the ECCS, whereas the problem described in the proposed Generic letter relates to "the adequacy of ECCS configuration control and

- operating practices." The actions requested by the proposed Generic Letter more appropriately relate to conduct and coordination of activities while in a shut down condition. By citing Section 50.46 as the basis for the compliance exception to the backfit provisions of 10 C.F.R. @50.109, the proposed Generic Letter 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. Accordmg 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."I' 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 design features of the ECCS may not be the most appropriate corrective actions to address this situation. The NRC Staff position implied in th 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.

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

WINSTON & STRMN

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U. S. Nuclear Regulaory Commission March 17.1997 Page 3 r-t Licensees have been made aware of the event discussed in the proposed G Letter by Information Notice 95-03 and other generic communications refere Gene _ric_ Letter. Licensees' actions taken as a result of these previous noti l- inadvertent operator actions that create a drain-down patl;way may be inspecte

. The information request is unnecessarily burdensome in scope and may requiie the ins .

or other extensive effort to generate the necessary information to respond. .The prop Letter would ask licensees, if the RCS is found to be susceptible to drain down event "each potential drain-down flow path (include piping sizes, identify flow path valves and normal positions, and identify valve interlocks and provisions'for valve position indication control room), describe potential valve testing manipulations or uses, and describ administrative drain-down events."

controls that are intended to be used to control valve manipulations t L

l We believe that this type of request is unwarranted and that the Staff has not show{

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that the burden on licensees is justified. In the Statement of Considerations for the revis Section 50.54(f), the NRC states that "[i]f extensive effort is reasonably anticipate be evaluated to determine whether the burden imposed by the information request i of the potential safety significance of the issue to be addressed. . . . Requests for informl l

determine compliance with existing facility requirements . . . usually are not made

{50.54(f). . . . The amendment of s50.54(f) should be read as indicating a strong .

of the Commission that extensive information requests be carefully scrutinized by s prior to initiating such requests. The Commission recognizes that there may be instances w is not clear whether a backfit will follow an information request. Those cases should be re in favor ofanalysis."F We believe this and the language of the rule itselfindicate the Comm original intent that Section 50.54(f) be used only for the most significant issues when the Commission must determine whether or not the license of a facility "should be modif 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 cou p

,__ inadvertently mispositioned during specific work activities on a case-by-case basis. Licensees been made aware of the importance ofproper administrative controls by the generic commun l

referenced in the proposed Generic Letter. We recommend that the NRC not issue the 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 add L

'l' 50 Fed. Reg. 38.112 (September 20.1985).

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- 10 C.F.R. s50.54(f).

w WINSTON & STRAWN

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U. S. Nuclear Regulaory 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.

x~ a-Sincerely, .

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1 Daniel F. Stenger Patricia L. Campbell Counsel for Nuclear Utility Backfitting and Reform Group W

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