Information Notice No. 95-54: Decay Heat Management Practices During Refueling Outages

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

                               December 1, 1995


NRC INFORMATION NOTICE 95-54:  DECAY HEAT MANAGEMENT PRACTICES DURING          
                               REFUELING OUTAGES


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 recent NRC assessments of licensee control of
refueling operations and the methods for removing decay heat produced from the
irradiated fuel stored in the spent fuel pool.  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.  

Background

The staff recently reviewed a design change and associated procedural controls
regarding spent fuel pool decay heat removal systems at Millstone Nuclear
Power Station, Unit 1, and full-core offloading controls at Cooper Nuclear
Station.  The staff evaluated overall controls on irradiated fuel movement and
the control of irradiated fuel decay heat removal during refueling operations,
including adequate adherence to final safety analysis report commitments,
implementation of procedures, procedural adequacy, and effectiveness of
training. 

Description of Circumstances

Millstone Unit 1

On October 18, 1993, the licensee for Millstone Unit 1 submitted Licensee
Event Report (LER) 93-11, in which it reported that it had determined through
engineering analysis that conditions may have existed during which the spent
fuel pool cooling system may have been incapable of maintaining spent fuel
pool temperature below the licensee's criteria of 66�C [150�F] design limit
for continued operation.  Specifically, the LER stated that (1) the licensee
had made inappropriate assumptions in the analysis performed in support of a
1988 spent fuel pool re-rack project, (2) the "normal" refueling sequence
described in the Millstone Unit 1 Updated Final Safety Analysis Report assumed
offload of only one third of a core, (3) a full-core offload considered in the
safety analysis report as an "emergency" (or abnormal discharge) offload was
normally performed at Millstone Unit 1, and (4) under certain circumstances

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Millstone Unit 1 may have operated outside its design basis for the spent fuel
pool.

The licensee recently implemented a modification to the shutdown cooling
system to provide additional spent fuel pool decay heat removal capability. 
In a July 28, 1995 submittal, the licensee stated that the modification would
enable Millstone Unit 1 to perform a full core discharge as a normal offload
practice.  Coincident with the development of the modification, the licensee
proposed a license amendment to impose technical specification controls on
shutdown cooling system operability, spent fuel pool temperature and decay
time prior to beginning offload activities.  In response to staff questions,
the licensee stated it had concluded during a review pursuant to 10 CFR 50.59,
that the proposed modification did not represent an unreviewed safety question
and as such did not require prior NRC approval, and that the license amendment
was not required but was being submitted to remove ambiguity regarding the
full core offload refueling practice.  

During its review of the procedural controls for the shutdown cooling-spent
fuel pool cooling cross-connect, the staff found that the administrative
procedures for the cross-connect, including controls for the cross-connect
valves and the spent fuel pool-reactor vessel weir gate were not sufficiently
explicit.  The licensee addressed these concerns.  Because the requested
specifications did not meet the criteria of 10 CFR 50.36 for inclusion as
limiting conditions of operation in the technical specifications, the staff
issued a license condition.  The license condition specifies that refueling
operations that include full core offload be conducted in accordance with the
revised controls proposed by the licensee.


Cooper Nuclear Station

On October 20, 1995, the operators of the Cooper nuclear station halted
movement of fuel from the reactor vessel to the spent fuel pool to perform a
review of the design and licensing basis and administrative controls
associated with the removal of decay heat from the spent fuel pool.  The
licensee concluded that no licensing restrictions regarding the practice of
conducting a full-core offload existed with regard to decay heat removal.  The
licensee further concluded that the installed spent fuel pool cooling system
and backup fuel pool cooling inter-tie from the residual heat removal system
had sufficient capacity to remove the decay heat from the irradiated spent
fuel and reactor cavity for postulated heat loads up to and including those
associated with a full-core offload.  

However, the licensee acknowledged that the description of the spent fuel pool
cooling system in the Cooper Updated Safety Analysis Report was confusing and
ambiguous.  Consequently, the licensee proposed revisions to that document to
clarify ambiguous language and performed a 10 CFR 50.59 analysis, which
documented the evaluation of the plant's licensing basis for the design and
operation of the spent fuel pool cooling system.  Upon approval of the changes
by the Station Operations Review Committee, the licensee updated its refueling
procedures to be consistent with the revised safety analysis report, and
proceeded with the full-core offload.
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Discussion

The functional capability to protect irradiated fuel from damage due to
inadequate decay heat removal is an important safety attribute.  Maintaining 
spent fuel pool water temperature below boiling temperature provides adequate
cooling for stored irradiated fuel.  However, prolonged operation at elevated
spent fuel pool temperatures may impair the capability of the purification
system to remove contaminants from the spent fuel pool coolant and increase
the rate of heat addition to the fuel storage area atmosphere to a value above
that assumed in the ventilation system design.  In addition, high spent fuel
pool temperatures may exceed the temperature used for thermal stress
computation in the structural analysis of the spent fuel pool liner and the
spent fuel pool structure itself.

Shutdown cooling systems, which are aligned to directly cool the reactor
vessel or reactor coolant system, are designed to remove the residual and
decay heat associated with the irradiated fuel in the reactor vessel in order
to bring the reactor coolant system to the cold shutdown condition.  Licensees
typically have procedures to maintain the removal of decay heat from the
vessel at all times while irradiated fuel is in the reactor vessel.

Similarly, systems are also installed in nuclear power plants to remove from
the spent fuel pool, decay heat generated by the stored irradiated fuel. 
However, these spent fuel pool cooling systems are designed with a lower heat
removal capacity relative to the shutdown cooling system based on the decrease
in decay heat generation within the irradiated fuel as the time after reactor
shutdown increases.  At some facilities, including most boiling water reactors
and some pressurized water reactors, the spent fuel pool cooling system is not
designed to remove the decay heat associated with a full core immediately
after shutdown and still maintain a bulk spent fuel pool temperature below
design-basis limits.  However, these facilities are designed with backup spent
fuel pool cooling systems, which are generally alternative operating modes of
the residual heat removal or shutdown cooling systems, that supplement the
spent fuel pool cooling system during periods shortly after reactor shutdown
when the decay heat load of a full core may exceed the heat removal capacity
of the normal spent fuel pool cooling system.

The capability of spent fuel pool cooling systems and backup spent fuel decay
heat removal systems is described in the Final Safety Analysis Report, as
updated for nuclear power plants.  The decay heat load scenarios used to
evaluate the adequacy of system heat rejection capability may be based on a
series of core offloads and an associated decay time.  These scenarios may be
described as "normal" or "abnormal" maximum heat loads for this purpose, which
is consistent with the NRC staff guidance for the review of spent fuel pool
cooling system design contained in Section 9.1.3 of the Standard Review Plan
(NUREG-0800).

Recent licensee reviews of refueling outage practices at Millstone Unit 1 and
Cooper found that the system design bases specified in Final Safety Analysis
Reports, as related to core offload practices, were ambiguous.  Administrative
controls on refueling outage plans and practices were inconsistent in regard  .                                                            IN 95-54
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to ensuring that temperature commitments for the spent fuel pool were
maintained through all phases of refueling operation.  Both licensees, after
clarifying and improving the design bases and administrative controls related
to refueling outages, determined that a routine practice of performing full
core offloads was acceptable.

The NRC has issued two information notices to alert licensees to potential
risks associated with a loss of spent fuel pool cooling.  NRC Information 
Notice 93-83, "Potential Loss of Spent Fuel Pool Cooling After a Loss-of-
Coolant Accident," was issued October 7, 1993, and described concerns found at
Susquehanna Steam Electric Station.  NRC Information Notice 93-83,  
Supplement 1, was issued August 8, 1995, to inform licensees of the results of
the NRC review of the concerns at Susquehanna.

The events described in this and previous information notices, and the plant
reviews discussed above, illustrate the importance of:

�     assuring that planned core offload evolutions, including refueling
      practices and irradiated decay heat removal, are consistent with the
      licensing basis, including the Final Safety Analysis Report, technical
      specifications, and license conditions;

�     assuring that changes are evaluated through the application of the
      provisions of 10 CFR Part 50.59, as appropriate; and

�     assuring that all relevant procedures associated with core offloads have
      been appropriately reviewed.

The staff is continuing to review this matter with respect to the need to
issue additional generic communications.

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.

                                    /s/'d by DMCrutchfield


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

Technical contacts:  Joseph W. Shea, NRR           Steven R. Jones, NRR
                     (301) 415-1428                (301) 415-2833

                     David L. Skeen, NRR
                     (301) 415-1174