United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 97-48: Inadequate or Inappropriate Interim Fire Protection Compensatory Measures

                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                         WASHINGTON, D.C.  20555-0001

                                 July 9, 1997


NRC INFORMATION NOTICE 97-48: INADEQUATE OR INAPPROPRIATE INTERIM FIRE
                              PROTECTION COMPENSATORY MEASURES


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 alert addressees to potential problems associated with the
implementation of interim compensatory measures for degraded or inoperable
plant fire protection features or degraded or inoperable conditions associated
with post-fire safe-shutdown capability.  It is expected that recipients will
review the information for applicability to their facilities and consider
actions, as appropriate, to avoid similar problems.  However, suggestions
contained in this information notice are not NRC requirements; therefore, no
specific action or written response is required.  

Description of Circumstances

The NRC foresaw cases in which fire protection features would be inoperable
and allowed licensees, through plant technical specifications (TS) or approved
fire protection plans controlled by license conditions, to implement
alternative actions to compensate for the inoperable condition or component
until permanent corrective actions are implemented.  In general, the
combination of appropriate compensatory measures and the defense-in-depth fire
protection features provides an adequate level of fire protection until the
licensees complete the corrective actions.  For common types of deficiencies
(e.g., an inoperable fire suppression system), the specific compensatory
measures are generally noted in either the TS or in the NRC-approved fire
protection program.  For unique plant-specific situations (e.g., inadequate
cable separation), the appropriate compensatory measures are determined by the
licensee on a case-by-case basis.  (This is discussed further later in the
information notice.)

NRC reviews of licensees' corrective actions to address degraded or inoperable
conditions associated with a plant's abilities to achieve and maintain post-
fire safe shutdown have noted some weaknesses in the interim compensatory
measures.  For example, some licensees have relied exclusively on the 1-hour
roving fire watch as an interim compensatory measure for deficiencies that
affect their abilities to achieve and maintain post-fire safe shutdown and 
have not considered or implemented other appropriate interim compensatory
measures, such as briefing operators on degraded post-fire, safe-shutdown-
system conditions; temporary repair procedures; temporary fire barriers; or
detection or suppression systems.  In addition,  

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NRC inspections of fire protection programs have found weaknesses in
licensees' fire watch training programs and in the conduct of fire watch
duties.  Specific examples documented in NRC inspection reports include the
following: 

.  During a November 1995 inspection at Waterford Generating Station
   (Inspection Report 50-382/95-020, Accession No. 9603250040), NRC inspectors
   found that the licensee's procedures redefined the intent of a continuous
   fire watch.  The licensee's procedure required a 15-minute roving fire
   watch, with a margin of 5 minutes, in lieu of a continuous fire watch.  In
   addition, required fire watches were not performed in areas in which fire
   detection equipment was out of service and in areas in which fire barriers
   were degraded.  This issue is currently under review by the NRC for
   disposition. 

.  During an April 1996 inspection at Palisades Nuclear Plant (Inspection
   Report 50-255/ 96-004, Accession No. 9605290105), the NRC inspector noted
   that the licensee relied on fire watches as compensatory measures for a
   number of post-fire, safe-shutdown-design deficiencies identified by the
   licensee pursuant to Appendix R to Part 50 of Title 10 of the Code of
   Federal Regulations (10 CFR Part 50) while long-term corrective actions
   were being evaluated.  The licensee did not alert the operators to these
   deficient conditions nor did it provide interim shutdown strategies to
   operators on how to successfully deal with potential fire damage to
   required safe-shutdown components.  The potential conse- quences could have
   been extensive since placing and maintaining the plant in post-fire hot
   standby could only have been achieved by significant operator actions,
   trouble- shooting, and repair activities to compensate for the post-fire,
   safe-shutdown-design deficiencies.  The sole use of a fire watch for a safe
   shutdown function which is not adequately protected against fire damage is
   an inappropriate application of a compensatory measure.  This problem
   resulted in a Severity Level III Violation.    

.  On October 17, 1996, Arkansas Nuclear One had an event involving a fire in
   a reactor coolant pump (RCP) and inadequacies associated with the RCP oil
   collection system.  The licensee's initial compensatory measures placed the
   lift oil pumps in the "stop" condition.  This action enhanced the fire
   protection annunciator procedures by giving additional guidance to the
   operators when responding to the RCP smoke detector alarm.  However, the
   licensee did not propose interim plant operational measures which would
   help operators identify and mitigate the consequence of potential oil
   leakage from the RCP oil system sites not protected by the RCP oil
   collection system.  This problem resulted in a Severity Level III
   Violation.

Discussion

In accordance with 10 CFR 50.48, "Fire protection," each operating nuclear
power plant must have a fire protection plan that satisfies Criterion 3 of
Appendix A to 10 CFR Part 50.  Licensees of plants licensed to operate before
January 1, 1979, are required to satisfy the provisions of Appendix R to 10
CFR Part 50, Section III.G, "Fire protection of safe shutdown 
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capability;" Section III.J, "Emergency lighting;" and Section III.O, "Oil
collection system for reactor coolant pump;" and the provisions of Appendix A
to the Branch Technical Position of the Auxiliary Power Conversion Systems
Branch (BTP APCSB) 9.5-1, "Guidelines for Fire Protection for Nuclear Power
Plants Docketed Prior to July 1, 1976."  Licensees of plants licensed to
operate after January 1, 1979, are required to satisfy the fire protection
provisions of their operating licenses.  Generally, these plants have fire
protection programs that satisfy the provisions of NUREG-0800, "Standard
Review Plan (SRP)," Section 9.5.1, "Fire Protection Program," July 1981.

Appendix A to BTP APCSB 9.5-1 specifies that licensees should establish
administrative controls for normal and abnormal conditions or such other
anticipated operations as modifications (e.g., breaking fire stops/penetration
seals, impairment of fire detection and suppression systems) and refueling
activities.  The controls should be reviewed by appropriate levels of licensee
management, and appropriate special actions and procedures, such as fire
watches or temporary fire barriers, should be implemented to ensure adequate
fire protection and reactor safety.  
 
On November 7, 1991, the NRC issued Generic Letter 91-18, "Information to
Licensees Regarding Two NRC Inspection Manual Sections on Resolution of
Degraded and Nonconforming Conditions and on Operability" (Accession No.
9111040293).  This generic letter contained two sections, "Resolution of
Degraded and Nonconforming Conditions" and "Operable/Operability:  Ensuring
the Functional Capability of a System or Component," to be included in Part
9900, "Technical Guidance," of the NRC Inspection Manual.  These additions are
based upon previously issued guidance. 

With respect to fire protection and post-fire, safe-shutdown capability, the
licensee may discover a previously unanalyzed condition.  Under these
conditions, Part 9900, "Technical Guidance:  Resolution of Degraded and
Nonconforming Conditions," Section 4.4, "Discovery of an Existing But
Previously Unanalyzed Condition or Accident," notes that the licensee, upon
discovering an existing but previously unanalyzed condition that significantly
compromises plant safety, shall report that condition in accordance with 10
CFR 50.72 and shall place the plant in a safe condition.  Once a degraded or
nonconforming condition of specific structures, systems, or components (SSCs)
is identified, an operability determination should be made as soon as
possible, consistent with the safety importance of the SSC affected.  For SSCs
that are outside plant TS, engineering judgment must be used to determine
safety significance.    

Part 9900, "Technical Guidance:  Resolution of Degraded and Nonconforming
Conditions," Section 4.6, "Reasonable Assurance of Safety," states that for
SSCs which are not expressly subject to TS and which are determined to be
inoperable, the licensee should assess the reasonable assurance of safety.  If
the assessment assures safety, then the facility may continue to operate while
prompt corrective action is taken.  As stated in Part 9900, the following
items are to be considered for such assessments:  the availability of
redundant or backup equipment, compensatory measures (including limited
administrative controls), and the conservatism and margins.

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


                                signed by S.H. Weiss for

                                Marylee M. Slosson, Acting Director
                                Division of Reactor Program Management
                                Office of Nuclear Reactor Regulation

Technical contacts:  Darrell Schrum, RIII
                     (630) 828-9741
                     E-mail:  dls3@nrc.gov

                     Rogelio Mendez, RIII
                     (630) 829-9745
                     E-mail:  rxm@nrc.gov

                     Patrick Madden, NRR
                     301-415-2854
                     E-mail:  pmm@nrc.gov
Page Last Reviewed/Updated Tuesday, December 03, 2013