ML17164A472

Discusses 416th Meeting of ACRS on 941208-10 Re Potential for Loss of Spent Fuel Pool Cooling Following loss-of-coolant Accident at Plant

ML17164A472
Person / Time
Site:	Susquehanna
Issue date:	12/19/1994
From:	Kress T NRC COMMISSION (OCM)
To:	Selin I, The Chairman NRC COMMISSION (OCM)
Shared Package
ML17164A471	List: ML17164A470 ML17164A472
References
NUDOCS 9412300203
Download: ML17164A472 (3)

Text

t UNITED STATES t NUCLEAR REGULATORY COMMISSION ADVISORYCOMMITTEEON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 December 19, 1994 The Honorable Ivan Selin Chairman U ~ S. Nuclear Regulatory Commission Washington, D.C.

20555-0001

SUBJECT:

LOSS OF SPENT FUEL POOL COOLING FOLLOWING A LOSS-OF-COOLANT ACCIDENT AT THE SUSQUEHANNA STEAM ELECTRIC STATION During the 416th meeting of the Advisory Committee on Reactor Safeguards, December 8-10,

1994, we discussed the NRC staff Draft Safety Evaluation Report

{DSER) dealing with the potential for loss of spent fuel pool cooling following a loss-of-coolant accident (LOCA) at the Pennsylvania Power and Light

{PP6L)

Company's Susquehanna Steam Electric Station Units 1

and 2.

During the

meeting, we had the benefit of discussions with representatives of the NRC staff, PP&L, and the individuals who brought this matter to the attention of the NRC on November 27,

1992, through a

10 CFR Part 21 notification.

We also had the benefit of the documents referenced.

We considered this matter previously during our May 5-7, 1994 meeting.

The 10 CFR Part 21 notification described the individuals'oncerns with:

(1) the ability of Susquehanna to provide adequate cooling of the spent fuel storage pool following various design-basis LOCAs; (2) the potential causes and consequences of failure to cool the spent fuel storage pool; and (3) numerous regulatory issues regarding potential design deficiencies.

The primary concern raised by the two individuals was a postulated failure to cool the spent fuel storage pool following a design-basis LOCA or a LOCA with a loss of offsite power (LOOP).

They posited that a design-basis LOCA would result in the failure of the nonsafety-related spent fuel pool cooling system.

They further posited that a design-basis LOCA results in the development of a TID 14844-like radiological source term inside the reactor building that would prevent operators from entering the building and restoring cooling to the spent fuel pool.

The individuals further postulated

that, upon boiling in the

pool,

. vapor would be transported throughout the reactor building by the ventilation systems and would eventually cause the failure of safety-related systems needed to mitigate the LOCA.

The ultimate consequences of 9'412300203 941223 PDR ADQCK 05000387 8

PDR

~,,

Enclosure

The Honorable Ivan Selin these boiling scenarios include severe core damage, failure of the stored spent fuel, and loss of primary and secondary containment.

The

DSER, which stands separate from the staff's regulatory compliance evaluation, includes a

review of certain specific aspects of the Susquehanna facility design and a deterministic examination of some of the physical phenomena involved.

The

" evaluation also includes a probabilistic analysis of postulated event sequences involving loss of the spent fuel storage pool cooling.

In our review of this matter, we were looking for answers to three questions:

1.

Is Susquehanna now operating without undue risk to the health and safety of the public?

2.

3.

Was Susquehanna operating in an unsafe condition prior to modifications and procedural changes that have been made?

Are there generic implications of undue risk at other operating plants?

Additionally, we have an interest in whether or not the postulated pool boiling sequences should have been part of the design-basis accident

and, thus, part of the licensing basis for Susquehanna.

Our interest here stems from our concerns about coherence in the regulatory process and about ill-advised actions that can create burdens on licensees without providing a corresponding increase in safety.

Clearly, the appropriate approach to answering the first question is to conduct a limited probabilistic risk assessment (PRA) for the plant as now configured, focusing on the LOCA sequences that can lead to spent fuel pool boiling.

The staff has done this and found that the core-damage frequency (CDF) is less than 1 x 10 '/yr.

This clearly indicates that the plant is not at undue risk from these particular sequences.

The appropriate approach to answering the second question is to repeat the limited PRA but with the plant in the as-found configuration before any modifications.

The staff has conducted this study and found that the risk was similarly low, with a CDF of 4 x 10-'/yr.

Our opinion on this issue rests on how well we think these PRAs were done and whether or not the results are credible.

Since we did not review these PRAs in any detail, we are unable at this time to make a judgment as to their quality.

Because the safety case rests primarily on the validity of the results of these

PRAs, we recommend that the PRAs and their associated uncertainty analyses

The Honorable Ivan Selin i

be given a thorough review.

The reviewers should pay particular attention to the treatment given the environmental effects brought about by LOCAs, including interfacing system LOCAs.

This area of PRA could use additional research by NRC.

We cannot judge the generic implications.

The low risk for the "as-found"'onfiguration (before modifications), indicated by the PRA result, indicates to us that spent fuel pool boiling is not likely to be of concern as a risk-contributor at other plants.

Nevertheless, we think it appropriate that NRC issue a generic notification to all licensees describing this particular issue and requesting a review of plant vulnerability to spent fuel pool boiling.

This could be an adjunct to the Individual Plant Examination (IPE) process.

With respect to the licensing-basis

issue, we have the following opinion.

If the PRA result indicating very low risk is correct, then it would be inappropriate at this time to consider augmenting the Susquehanna licensing basis with the postulated pool-boiling sequences.

Sincerely,

~ s.

/'.

S. Kress Chairman

References:

1.

Letter dated October 24,

1994, from Gary M. Holahan, Office of Nuclear Reactor Reguation, NRC, to J.

T. Larkins, Executive

Director, ACRS,

Subject:

409th ACRS Meeting Followup Matters and transmitting Draft Safety Evaluation Report 2.

Letter dated May 16,

1994, from D. Lochbaum and D. Prevatte, Members of Public, to J. T. Larkins, Executive Director, ACRS,

Subject:

Susquehanna Steam Electric Station Units 1 and 2

Loss of Spent Fuel Pool Cooling Licensing Basis