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83

Potential Loss of Spent Fuel Pool Cooling after a Loss­Of­Coolant

Accident or a Loss of Offsite Power

UNITED STATES

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

(LOCA)

Addressees

All holders of operating licenses or construction permits for boiling water

reactors (BWR).

Purpose

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

the concerns.

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

information.

Discussion

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
9310070169. IN 93‐83

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

transfer system.

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

buildings.

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

sink.

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October 7, 1993

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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

BWRs.

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

(301) 504‐2860

George Hubbard, NRR

(301) 504‐2870

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