Information Notice No. 92-36, Supplement 1: Intersystem LOCA Outside Containment

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                            WASHINGTON, D.C.  20555

                               February 22, 1994


NRC INFORMATION NOTICE 92-36, SUPPLEMENT 1:  INTERSYSTEM LOCA OUTSIDE
                                             CONTAINMENT


Addressees

All holders of operating licenses or construction permits for nuclear power
reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to inform addressees about technical findings resulting from the NRC
program for resolving Generic Safety Issue (GSI) 105, "Interfacing System
Loss-of-Coolant Accidents [LOCAs] at Light Water Reactors."  It is expected
that recipients will review the information for applicability to their
facilities and consider actions as appropriate.  However, suggestions
contained in this information notice are not NRC requirements; therefore, no
specific action or written response is required.

This information may help licensees in their plant evaluations recommended by
Generic Letter 88-20, "Individual Plant Examination for Severe Accident
Vulnerabilities - 10 CFR 50.54(f)," November 23, 1988.  As discussed below,
the NRC staff considers that participation in the individual plant examination
program is sufficient to resolve GSI 105.

Background

In NRC Information Notice 92-36, the staff stated that the interfacing system
loss-of-coolant accidents (ISLOCAs) of most concern are those accidents during
which the break flow is discharged outside the reactor containment building.
The two main reasons for this concern are:  (1) potential high offsite
radiological consequences caused by radioactive effluent bypassing the
containment and (2) potential loss of long-term core cooling resulting from
loss of reactor coolant system (RCS) inventory that would otherwise be
available for recirculation from the containment sumps.

Several draft plant-specific probabilistic risk assessments (PRAs) of core
damage frequency from an ISLOCA were cited in Information Notice 92-36.  The
PRAs have since been completed (Refs. 1-4).

The major impetus to the ISLOCA PRA program was the fact that for ISLOCA
precursor events human error was a major contributor, a situation not
predicted by PRAs at the time because of insufficient modeling of human error.

9402150320.

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The importance of human error prompted the NRC staff to question early PRA
results.  The staff considered that (1) the frequencies of ISLOCA precursor
events that were strongly influenced by human error were higher than expected;
(2) because of the potential for radioactive effluent bypassing the
containment, some ISLOCA sequences could result in early release with high
radiological consequences; and (3) additional PRAs were needed that accounted
for the influences of human error and human response.

Discussion

The staff presented some observations in Tables 1 and 2 of Information
Notice 92-36.  These observations were based on root cause analyses of ISLOCA
precursor events; extensive plant inspections; and detailed, but at that time
incomplete, analyses of a sample of pressurized water reactors.  These
observations remain essentially unchanged.

The completed ISLOCA PRAs have provided the staff with useful insights in
addition to those given in Table 2 of Information Notice 92-36:

    -   The susceptibility to an ISLOCA is highly plant specific.  ISLOCA
        contributors important at one plant are not necessarily important at
        another.

    -   Operator errors can be important contributors.  Important operator
        errors include valve alignment errors during transitions between
        operating modes.

    -   Recovery actions can be more important than indicated in previous
        PRAs.  Rapid isolation of the break can prevent the compounding of
        potential problems with emergency core cooling equipment and borated
        water supplies.

    -   Likely locations of breaks in low-pressure systems, given ISLOCA
        initiations, are not readily predetermined and may require some
        analysis.  ISLOCA training programs and simulations at some units may
        not include the most likely and potentially most severe break
        locations.

    -   Emergency operating procedures at some plants do not allow attention
        to be focused on break isolation until actions are completed that
        ensure injection equipment is operating properly.  For large breaks,
        this could be inefficient use of time if some emergency core cooling
        equipment is susceptible to auxiliary building flooding or other
        damage from break effluent.

    -   The harsh environment of an ISLOCA may prohibit personnel access to
        remote stations.  This may complicate long-term cooling and hinder
        efforts to stop the loss of coolant outside the containment.
.

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    -   Symptom-based procedures may lead the operator to realize that an
        ISLOCA has occurred.  However, unless the emergency procedures refer
        to plant provisions for conserving and replenishing water, the
        operator may have difficulty managing the accident.  If the emergency
        core cooling system water is not replenished, an ISLOCA may lead to
        core damage, even after the leak has been isolated.

The GSI 105 resolution program also produced information useful in determining
likely locations of breaks in low-pressure systems (Refs. 7 and 8).

Resolution of Generic Safety Issue 105

The NRC staff gave this issue a high priority ranking and pursued its
resolution.  In resolving the issue, the staff conducted analyses of units
representative of all nuclear steam supply system vendors and considered human
errors, component fragilities and the survivability of recovery equipment
(References 1-6), concluding that the units studied posed little risk from
ISLOCA.  The study of ISLOCA at a boiling water reactor confirmed past
studies, which indicated little risk contribution from ISLOCA sequences.  The
staff found that ISLOCAs at pressurized water reactors were plant-specific in
nature but noted that the ongoing IPE program includes licensee analysis of
ISLOCA sequences.  Thus, in the resolution of GSI 105, it was recognized that
licensees and applicants, in response to Generic Letter 88-20, were already
performing IPEs to identify and take steps to prevent or mitigate severe
accident sequences, including ISLOCAs, at their plants.  This ongoing IPE
program together with the relatively small core damage frequency contribution
from ISLOCA identified in the completed PRAs, resulted in the conclusion
(References 7 and 8) that participation in the IPE program was sufficient to
resolve GSI 105.  Thus, this issue was resolved and no new requirements were
established, as documented in NUREG-0933, "A Prioritization of Generic Safety
Issues," as issued June 30, 1993.

Information in References 1-6 contain the latest ISLOCA analyses by the NRC
staff and contractors.  Likely areas of greatest interest to IPE analysts are
the following:

    -   pressure isolation valve initiating events, (hardware faults, human
        errors, test and maintenance procedures or combinations of these
        items)

    -   pressure-induced failure or rupture of the interfacing system

    -   rupture detection and diagnosis

    -   isolation of the rupture

    -   mitigation of accident sequences

.

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Reference 1 lists significant items in each of these areas and presents a
screening procedure for rapidly identifying gross ISLOCA vulnerabilities,
using numerical scores based on values assigned to the status of significant
items in these areas.

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please call one
of the technical contacts listed below or the appropriate Office of Nuclear
Reactor Regulation (NRR) project manager.


                                    ORIGINAL SIGNED BY


                                    Brian K. Grimes, Director
                                    Division of Operating Reactor Support
                                    Office of Nuclear Reactor Regulation

Technical contacts:  Gary Burdick, RES
                     (301) 492-3812

                     Vern Hodge, NRR
                     (301) 504-1861

Attachments:
1.  List of References
2.  List of Recently Issued NRC Information Notices
.

                                                      Attachment 1
                                                      IN 92-36, Supp. 1
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                                                      Page 1 of 1

                              LIST OF REFERENCES



1.  W. J. Galyean et al., "ISLOCA Research Program Final Report,"
    NUREG/CR-5928, July 1993.

2.  W. J. Galyean and D. I. Gertman, "Assessment of ISLOCA Risk - Methodology
    and Application to a Babcock and Wilcox Nuclear Power Station,"
    NUREG/CR-5604, May 1992.

3.  D. L. Kelly, J. L. Auflick, and L. N. Haney, "Assessment of ISLOCA Risks -
    Methodology and Application to a Westinghouse Four Loop Ice Condenser
    Plant," NUREG/CR-5744, May 1992.

4.  D. L. Kelly, J. L. Auflick, and L. N. Haney, "Assessment of ISLOCA Risks:
    Methodology and Application to a Combustion Engineering Plant,"
    NUREG/CR-5745, May 1992.

5.  D. A. Wesley et al., "Pressure-Dependent Fragilities for Piping
    Components," NUREG/CR-5603, October 1990.

6.  D. A. Wesley, "Screening Methods for Developing Internal Pressure
    Capacities for Components in Systems Interfacing With Nuclear Power Plant
    Reactor Coolant Systems," NUREG/CR-5862, May 1992.

7.  "Regulatory Analysis for the Resolution of Generic Issue 105: Interfacing
    System Loss of Coolant Accident in Light Water Reactors," NUREG-1463,
    July 1993.

8.  Memorandum for Eric S. Beckjord from James M. Taylor, "Technical
    Resolution of Generic Issue 105 (GI-105) - ISLOCA in LWRS,"
    June 21, 1993.