NRC: Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions (Generic Letter 96-06, Supplement 1)

file:///X|/...Operability%20and%20Containment%20Integrity%20During%20Design-Basis%20Accident%20Conditions%20(Generic%20Letter%2096-06,%20Supplement%201).html[4/20/2015 2:02:33 PM]

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Assurance of Equipment Operability and Containment Integrity During

Design-Basis Accident Conditions (Generic Letter 96-06, Supplement 1)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001

November 13, 1997

NRC GENERIC LETTER 96-06,

SUPPLEMENT 1:

ASSURANCE OF EQUIPMENT OPERABILITY AND CONTAINMENT INTEGRITY DURING DESIGN-BASIS

ACCIDENT CONDITIONS

Addressees

All holders of operating licenses for nuclear power reactors 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 supplement to Generic Letter (GL) 96-06, "Assurance of Equipment Operability and

Contain-ment Integrity During Design-Basis Accident Conditions," to inform addressees about ongoing efforts and new developments associated with

GL 96-06 and to provide additional guidance for completing corrective actions. Addressees may find this information useful in planning and scheduling

future actions associated with GL 96-06. This generic letter supplement contains no new NRC requirements. Furthermore, no specific action or written

response is required.

Background

GL 96-06 was issued on September 30, 1996, to address the following issues of concern:

1. Cooling water systems serving the containment air coolers may be exposed to the hydrodynamic effects of waterhammer during either a loss-ofcoolant accident (LOCA) or a main steam line break (MSLB). These cooling water systems were not designed to withstand the hydrodynamic

effects of waterhammer and actions may be needed to satisfy system design and operability requirements.

2. Cooling water systems serving the containment air coolers may experience two-phase flow conditions during postulated LOCA and MSLB

scenarios. The heat removal assumptions for design-basis accident scenarios are based on single-phase flow conditions and actions may be

needed to satisfy system design and operability requirements.

3. Thermally induced overpressurization of isolated water-filled piping sections in containment could jeopardize the ability of accident-mitigating

systems to perform their safety functions and could lead to a breach of containment integrity through bypass leakage. Actions may be needed to

satisfy system operability requirements.

9711050091

GL 96-06, Supplement 1

November 13, 1997

GL 96-06 states--

"If systems are found to be susceptible to the conditions discussed in this generic letter, addressees are expected to assess the operability of affected

systems and take corrective action as appropriate in accordance with the requirements stated in 10 CFR Part 50 Appendix B and as required by the

plant Technical Specifications.

GL 96-06 refers to GL 91-18, "Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming

Conditions and on Operability," for guidance on the resolution of issues identified in GL 96-06. (Note: GL 91-18, Revision 1 was issued on October 8,

1997, to inform licensees of the issuance of a revised section of the NRC Inspection Manual, Part 9900, "Technical Guidance," on the resolution of

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NRC: Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions (Generic Letter 96-06, Supplement 1)

file:///X|/...Operability%20and%20Containment%20Integrity%20During%20Design-Basis%20Accident%20Conditions%20(Generic%20Letter%2096-06,%20Supplement%201).html[4/20/2015 2:02:33 PM]

degraded and nonconforming conditions. The GL 96-06 reference to GL 91-18 remains valid since it refers only to the "Technical Guidance" on

operability, which was not changed by the issuance of GL 91-18, Revision 1.) Criterion XVI, "Corrective Actions," of Appendix B to 10 CFR Part 50

states, in part, "Measures shall be established to assure that...nonconformances are promptly identified and corrected." In this regard, GL 91-18 states

that the timeliness of corrective actions should be commensurate with the safety significance of the issue, and that the corrective action requirements

of Appendix B may be satisfied by making changes in the design of the plant in lieu of restoring the affected equipment to its original design. In one

example, GL 91-18 specifically discusses the use of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section III, Appendix F, criteria for interim operability determinations for degraded and nonconforming piping and pipe supports. It states that the use

of Appendix F criteria is valid until the next refueling outage when the supports are to be restored to the final safety analysis report criteria.

Addressees have responded to the generic letter and have established schedules for resolving the GL 96-06 issues. The NRC staff is currently reviewing

the information that has been submitted.

Discussion

Implementing corrective actions to resolve the GL 96-06 issues can have a significant impact on outage schedules and resources, and some

addressees have indicated that it would be prudent to take more time to better understand the specific concerns that have been identified in order to

optimize whatever modifications are needed and to assure that they do not ultimately result in a detriment to safety. Current issues and ongoing

efforts that could influence an addressee's decision in planning corrective actions include: (1) risk implications of installing relief valves to deal with the

thermal overpressurization issue; (2) feasibility of using the acceptance criteria contained in Appendix F to Section III of the American Society of

Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for the permanent resolution of the GL 96-06 issues; (3) ongoing tests by the

Electric Power Research Institute to support a generic resolution of the overpressurization of piping issue; and (4) questions regarding the staff's

closure of Generic Safety Issue 150, "Overpressurization of Containment Penetrations." Risk insights and industry initiatives that are being considered

or that may be proposed could also influence the course of action that addressees take to resolve the GL 96-06 issues.

Addressees are responsible for assessing equipment operability, determining actions, and establishing schedules that are appropriate for resolving the

specific conditions that have been identified. In determining the appropriate actions and schedules for resolving GL 96-06 issues, addressees should

consider, for example, the continued validity of existing operability determinations, compensatory actions required to maintain operability, the safety

significance associated with the specific nonconformances or degraded conditions that have been identified, risk insights, and the time required to

complete any generic initiatives and/or plant-specific actions (e.g., engineering evaluations, design change packages, material procurement, and

equipment modification and installation). Also, analytical solutions employing the permanent use of the acceptance criteria contained in the ASME

Code, Section III, Appendix F (or other acceptance criteria) may present viable alternatives to plant modifications and can be used where appropriate,

justified, and evaluated in accordance with NRC requirements such as 10 CFR 50.59, as applicable. Addressees may find the revised guidance contained

in GL 91-18, Revision 1, dated October 8, 1997, helpful in determining appropriate actions and schedules. Although adjustments in schedules may be

warranted on the basis of these (and other) considerations, specific actions that have been defined and are clearly needed should not be delayed

without suitable justification.

It is the staff's current position that addressees can use the ASME Code, Section III, Appendix F criteria for interim operability determinations for

degraded and nonconforming piping and pipe supports until permanent actions have been identified and approved by the NRC (as applicable) for

resolving the GL 96-06 issues. This guidance supplements the guidance provided by GL 91-18 for resolution of the GL 96-06 issues.

In order to further facilitate resolution of the GL 96-06 issues, the NRC will participate in a public workshop scheduled for December 4, 1997. The

workshop proceedings will be summarized by the NRC staff and made publicly available. The need for additional NRC guidance and generic

communication will be considered upon completion of the workshop.

Requested Information

Addressees who choose to revise their commitments for resolving the GL 96-06 issues should submit a revised response to the generic letter. Revised

responses should include appropriate discussion of the considerations discussed above, the current resolution status and actions remaining to be

completed, and plans being considered for final resolution of the GL 96-06 issues.

Federal Register Notification

Because this GL supplement is informational, requires no specific action or response, and is the result of ongoing efforts between NRC staff and

addressees to resolve GL 96-06 issues, there is no need for additional opportunities for comment. Accordingly, a notice of opportunity for public

comment was not published in the Federal Register. However, comments on the content of this supplement to GL 96-06 may be sent to the U.S.

Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-001.

Paperwork Reduction Act Statement

For those addressees who find it necessary to revise their commitments for resolving the GL 96-06 issues, this generic letter supplement contains

information collections that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved

by the Office of Management and Budget, approval number 3150-0011, which expires on September 30, 2000.

The public reporting burden for addressees who find it necessary to revise their response to GL 96-06 is estimated to average 40 hours per response,

including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and

reviewing the collection of information. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the collection of

information contained in the generic letter and on the following issues:

(1) Is the proposed collection of information necessary for the proper performance of the functions of the NRC, including whether the information will

have practical utility?

(2) Is the estimate of burden accurate?

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

(4) How can the burden of the collection of information be minimized, including the use of automated collection techniques?
NRC: Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions (Generic Letter 96-06, Supplement 1)

file:///X|/...Operability%20and%20Containment%20Integrity%20During%20Design-Basis%20Accident%20Conditions%20(Generic%20Letter%2096-06,%20Supplement%201).html[4/20/2015 2:02:33 PM]

Send comments on any aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records

Management Branch, T-6F33, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001, and to the Desk Officer, Office of Information and

Regulatory Affairs, NEOB-10202 (3150-0011), Office of Management and Budget, Washington, D.C. 20503.

If you have any questions about this matter, please contact the lead project manager or one of the technical contacts listed below, or the appropriate

Office of Nuclear Reactor Regulation (NRR) project manager for a specific nuclear power plant.

signed by D.B. Matthews for

Jack W. Roe, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: James Tatum, NRR

301-415-2805

E-mail: jet1@nrc.gov

John Fair, NRR

301-415-2759

E-mail: jrf@nrc.gov

Lead Project Manager: Beth Wetzel, NRR

301-415-1355

E-mail: baw@nrc.gov

Page Last Reviewed/Updated Friday, June 28, 2013

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