United States Nuclear Regulatory Commission - Protecting People and the Environment

Information Notice No. 93-83, Supplement 1: Potential Loss of Spent Fuel Pool Cooling after a Loss-Of-Coolant Accident or a Loss of Offsite Power

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

                                August 24, 1995


NRC INFORMATION NOTICE 93-83, SUPPLEMENT 1:  POTENTIAL LOSS OF SPENT FUEL POOL
                                             COOLING AFTER A LOSS-OF-COOLANT
                                             ACCIDENT OR A LOSS OF OFFSITE
                                             POWER


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 NRC staff findings regarding the risk associated
with the potential loss of spent fuel pool (SFP) cooling.  It is expected that
recipients will review this information notice for applicability to their
facilities and consider any appropriate actions.  However, suggestions
contained in this information notice are not NRC requirements; therefore, no
specific action or written response is required.

Background

The staff has been evaluating a report made under Part 21, "Reporting of
Defects and Noncompliance," of Title 10 of the Code of Federal Regulations  
(10 CFR), which two engineers, who formerly worked under contract for the 
Pennsylvania Power and Light Company, filed on November 27, 1992.  In the
report, the two engineers contended that the design of the Susquehanna Steam
Electric Station (SSES) failed to meet numerous regulatory requirements with
respect to a postulated sustained loss of the cooling function for the SFP
that mechanistically results from a loss-of-coolant accident (LOCA) or a loss
of offsite power (LOOP).  The report provided a series of detailed technical
and regulatory arguments to support this assertion.  It also postulated that
subsequent boiling of the SFP would cause failure of equipment necessary to
mitigate the accident or to safely reach a shutdown condition because of the
adverse environmental conditions created by SFP boiling within the reactor
building.  As a result of these equipment failures, severe offsite
consequences would result. 

Units 1 and 2 at SSES are boiling water reactors with Mark II containments
designed by General Electric Company.  The SFP and associated systems for each
unit are located in each unit's reactor building.  The surface of the SFPs is
on the common refueling floor, which spans the uppermost level of the two
reactor buildings.  The two SFPs communicate through a common cask storage pit
when the path is not isolated by gates.  The SFP cooling system for each unit
at the SSES consists of three parallel heat exchangers and three pumps.  Water

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to make up for evaporation and small leakage losses from the SFP is normally
supplied by the condensate transfer system.

The normal SFP cooling system and the normal system used for adding SFP makeup
water are not designed to remain functional after design-basis events.
However, the residual heat removal (RHR) system is designed to operate after
these events and can be aligned to cool the SFP by manual operation of valves
in the reactor building.  The emergency service water system is also designed
to operate after these events and can be aligned to provide water to the SFP
to make up for evaporative losses by manual operation of valves in the reactor
building.

Description of Circumstances
                       
The staff completed an assessment of safety with regard to a loss of SFP
cooling and determined that the concerns identified in the Part 21 report were
of low safety significance for SSES.  The assessment included an engineering
evaluation of the capability to recover from or mitigate a loss of SFP
cooling, and a quantitative estimation of the frequency of a sustained loss of
SFP cooling based on the findings of the engineering evaluation.  This
assessment is documented in a final safety evaluation report, which is
available for public review.  The staff considered comments on the draft
safety evaluation report from the authors of the Part 21 report, from
Pennsylvania Power and Light Company (the licensee for SSES), and from the
Advisory Committee on Reactor Safeguards for inclusion in the final safety
evaluation report.  The report was issued to Pennsylvania Power and Light
Company, Docket Nos. 50-387 and 50-388, on June 19, 1995.

While the staff was evaluating the Part 21 report, the licensee for SSES
initiated several actions to improve the capability to recover from a loss of
SFP cooling.  These actions included the following:  (1) committing to operate
with the two SFPs cross-connected through the cask pit to increase the
redundancy of cooling systems for the combined SFPs; (2) committing to conduct
testing and analyses that support assumptions regarding the reliability of the
SFP cooling assist mode of the RHR system; (3) completing analyses that
support modifications and procedural changes; (4) completing installation of
instrumentation to improve the capability to monitor SFP conditions; and 
(5) completing changes to off-normal and emergency procedures that improve the
reliability of recovery from a loss-of-SFP-cooling event.

The staff used both deterministic and probabilistic safety assessment
techniques to evaluate the safety implications of events involving a loss of
SFP cooling.  Because the staff did not consider a detailed evaluation of the
effects of SFP boiling necessary, based on an initial assessment of risk, the
staff elected to quantitatively estimate the frequency of SFP boiling and base
decisions regarding further evaluations on that estimate.

The staff's deterministic engineering evaluation of the capability to recover
from or mitigate a loss of SFP cooling identified important features of SSES
for modeling in the probabilistic safety assessment.  These characteristics
included the following:  (1) on the basis of licensee commitments and outage
management procedures, the time to the onset of pool boiling after a loss of.                                                            IN 93-83, Supp. 1
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cooling will exceed 25 hours; (2) natural circulation flow will maintain the
temperature difference between the two pools less than 17�C [30�F] with the
pools cross-connected through the common cask pit, thereby allowing a single
fuel pool cooling system of adequate capacity aligned to either pool to
prevent boiling in both pools; (3) equipment failures and human errors, which
are explicitly modeled in the safety assessment, are the significant failure
modes for the normal SFP cooling system; and (4) the SFP cooling assist mode
of the RHR system will provide a reliable means of cooling one or both pools
when access to the reactor building for manual system alignment is available.

The safety assessment quantitatively estimated the frequency of reaching a
near-boiling condition, which could add significant heat and water vapor to
the reactor building atmosphere, on the basis of the above information.  The
staff estimated that the actions the licensee has implemented to improve the
capability to recover from a loss of SFP cooling have reduced the near-boiling
frequency from 7.0E-5 per plant-year to 2.0E-5 per plant-year.

The dominant sequences for near-boiling frequency involve an extended LOOP,
but sequences involving a LOCA or a shorter LOOP are also significant.  The
dominance of sequences involving a LOOP reflects the reliance of the normal
SFP cooling system on offsite sources of electrical power and the limited
availability of the RHR system for fuel pool cooling because of the RHR
system's primary reactor vessel decay heat removal function.  Sequences
involving a LOCA were identified as significant specifically because the RHR
system in the affected unit is assumed to be unavailable for fuel pool
cooling. 

Despite the relatively small fraction of an operating cycle that each unit at
SSES was assumed to be in a refueling outage, the sequences occurring during
refueling outage periods that were examined dominated the near-boiling
frequency.  Two factors contributed to this result:  the relatively shorter
time to reach boiling after a loss of SFP cooling because of the practice of
conducting full-core off-loads at SSES and the practice of removing systems
associated with the outage unit that contribute to SFP decay heat removal
capability from service for maintenance during refueling outages.

To address generic concerns identified through the Part 21 report and separate
concerns related to spent fuel storage pools identified during a special
inspection at a permanently shutdown reactor facility (see NRC Bulletin 94-01,
"Potential Fuel Pool Draindown Caused by Inadequate Practices at Dresden 
Unit 1," dated April 14, 1994), the staff has developed and begun implementing
a generic action plan.  The generic plan includes the following actions: 
(1) a search and analysis of information regarding spent fuel storage pool
issues, (2) an assessment of the operation and design of spent fuel storage
pools at selected reactor facilities, (3) an evaluation of the assessment
findings for safety concerns, and (4) selection and execution of an
appropriate course of action based on the safety significance of the findings. 
During these assessments, the staff will examine those features that were
identified at SSES as important to the acceptably low level of risk from loss-
of-SFP-cooling events.
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Discussion

The functional capability to protect the reactor coolant pressure boundary, to
mitigate the effects of potential design-basis events, and to shut down the
reactor and maintain it in a safe shutdown condition are important safety
attributes.  Nuclear power plants are designed so that the potential for loss
of the capability to perform any of these functions is remote.  Adverse
environmental conditions, which may affect many components simultaneously,
have the potential to disable the redundant equipment that provides this
capability.

The staff conducted a licensing-basis review for SSES, which is documented in
Appendix A to the final safety evaluation report, and concluded that a loss of
SFP cooling initiated by a seismic event (seismically induced LOOP) was
considered in originally granting the facility's license.  The staff concluded
that, with the exception of seismically induced design-basis events, the
development of an adverse environment in the reactor building as a result of a
loss of SFP cooling is outside the licensing basis for SSES.  However, it also
concluded that the licensing basis with regard to SFP cooling at other
facilities may vary widely from that of SSES.  Therefore, the conclusion that
the development of an adverse environment in the reactor building as a result
of a loss of SFP cooling is outside the licensing basis at SSES may not be
valid at other facilities.

The staff performed a safety assessment to evaluate the frequency of near-
boiling events in the SFPs at SSES and found that the potential for such an
event was acceptably remote at SSES.  After analyzing the safety assessment
results, the staff concluded that the potential for reaching a near-boiling
condition in the SFP was remote principally because of the diverse installed
systems available for fuel pool cooling and the administrative controls that
ensured an extended period for recovery of cooling before the onset of
boiling. 
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This information notice requires no specific action or written response.  
If you have any questions regarding 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.

                                   /s/'d by DMCrutchfield


                                   Dennis M. Crutchfield, Director
                                   Division of Reactor Program Management
                                   Office of Nuclear Reactor Regulation

Technical contacts:  Steven Jones, NRR
                     (301) 415-2833

                     Joseph Shea, NRR
                     (301) 415-1428

                     David Skeen, NRR
                     (301) 415-1174
Page Last Reviewed/Updated Tuesday, November 12, 2013