Potential Loss of Spent Fuel Pool Cooling after a Loss-Of-Coolant Accident or a Loss of Offsite Power
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555
October 7, 1993
Information Notice No. 93-83: POTENTIAL LOSS OF SPENT FUEL POOL COOLING
FOLLOWING A LOSS OF COOLANT ACCIDENT
All holders of operating licenses or construction permits for boiling water
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to an issue the NRC is evaluating concerning the
potential loss of spent fuel pool (SFP) cooling following a LOCA. It is
expected that recipients will review the information 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.
Description of Circumstances
On November 27, 1992, a 10 CFR Part 21 notification was filed to notify the
NRC of concerns regarding the potential effects of a loss of SFP cooling
coincident with a LOCA at Susquehanna Steam Electric Station (SSES). Since
the initial submittal, additional submittals dated December 14, 1992, and
January 2, March 31, August 13, and October 1, 1993, have been made regarding
In response to these concerns, Pennsylvania Power and Light Company, the
licensee for SSES, has made submittals to the NRC dated May 24, July 6, and
August 16, 1993. The licensee met with the NRC on March 18 and July 8, 1993.
The NRC is currently evaluating the 10 CFR Part 21 notification and subsequent
Units 1 and 2 at SSES are BWRs with Mark II containments designed by the
General Electric Company. The SFPs for each unit are located above each
reactor in a reactor building common area. The two SFPs communicate through a
common cask storage pit when the path is not isolated by gates. The SFP
cooling systems for Units 1 and 2, as described in the updated final safety
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October 7, 1993
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analysis report (UFSAR), are non-seismic Category I, non-Class 1E powered, and
Quality Group C systems. The SFP cooling system for each unit consists of
three parallel heat exchangers, cooled by non-Class 1E service water, and
three pumps. During normal operation, the water temperature of the SFP is
maintained below 52�C [125�F]. Makeup water to accommodate for evaporation
and small leakage losses from the SFP is normally supplied by the condensate
During refueling outages, the residual heat removal (RHR) system is designed
to provide supplemental cooling to the SFP. The RHR system is connected to
the SFP by manually operating valves in the reactor building. The RHR system
cools the SFP using seismic Category I piping and can be isolated from the
non-seismic SFP cooling systems. The seismic Category I emergency service
water system also is available to provide makeup water for evaporative losses.
This system also requires the operation of manual valves in the pool area.
A LOCA coincident with a loss of SFP cooling could potentially limit recovery
actions. A LOCA in one unit may restrict access to the reactor building for
that unit. The transfer of steam or radioactive materials through the
heating, ventilation, and air conditioning systems also may restrict access to
the adjacent reactor building. Because entry to the reactor building is
necessary to provide a method of SFP cooling or makeup water addition when the
normal SFP cooling and make-up systems are inoperable, a delay in accessing
the reactor building may result in the SFP water boiling.
The submitted information identified the following concerns:
Potential loss of normal SFP cooling and makeup water systems as a result of
piping stresses caused by LOCA-induced hydrodynamic effects in the reactor
Potential inability to align emergency methods of SFP cooling and makeup water
addition under post-LOCA conditions.
Potential loss of safety-related equipment as a result of the temperature and
steam effects of SFP water boiling within the reactor building.
Potential loss of safety-related equipment as a result of flooding from
condensation of water vapor created by boiling the SFP water.
Adequacy of instrumentation to monitor SFP temperature and level.
Acceptability of the source term used to predict accessibility to the SFP area
and reactor building.
Consideration of SFP heat loads in the design basis for the ultimate heat
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Consideration that a loss of offsite power may last longer than 24 hours.
The NRC staff is evaluating these concerns and the licensee's actions as they
relate to the safe operation of SSES. The NRC staff also is evaluating the
safety significance of the concerns and their generic applicability to other
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 project manager.
/S/'D BY AECHAFFEE FOR/
Brian K. Grimes, Director
Division of Operating Reactor Support
Office of Nuclear Reactor Regulation
Technical contacts: David H. Shum, NRR
George Hubbard, NRR
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