Information Notice 1993-83, Potential Loss of Spent Fuel Pool Cooling After a Loss-of-Coolant Accident or a Loss of Offsite Power

Potential Loss of Spent Fuel Pool Cooling After a Loss-of-Coolant Accident or a Loss of Offsite Power

ML031070131
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	08/24/1995
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:
References
IN-93-083, Suppl 1, NUDOCS 9508180256
Download: ML031070131 (12)

V

I

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

9508180256 X

5 Ath

IN 93-83, Supp. 1 August 24, 1995 the SFP is normally

to make up for evaporation and small leakage losses from

supplied by the condensate transfer system.

for adding SFP makeup

The normal SFP cooling system and the normal system used events.

water are not designed to remain functional after design-basis

designed to operate after

However, the residual heat removal (RHR) system is operation of valves

these events and can be aligned to cool the SFP by manual

water system is also designed

in the reactor building. The emergency service water to the SFP

to operate after these events and can be aligned to provide

operation of valves in the reactor

to make up for evaporative losses by manual

building.

Description of Circumstances

a loss of SFP

The staff completed an assessment of safety with regard to Part 21 report were

cooling and determined that the concerns identified in the

an engineering

of low safety significance for SSES. The assessment included loss of SFP

from or mitigate a

evaluation of the capability to recover sustained loss of

the frequency of a

cooling, and a quantitative estimation of This

SFP cooling based on the findings of the engineering evaluation.

which is

assessment is documented in a final safety evaluation report,on the draft

available for public review. The staff considered comments

report, from

safety evaluation report from the authors of the Part 21 and from the

Pennsylvania Power and Light Company (the licensee for SSES),

for inclusion in the final safety

Advisory Committee on Reactor Safeguards Power and Light

evaluation report. The report was issued to Pennsylvania

Company, Docket Nos. 50-387 and 50-388, on June 19, 1995.

for SSES

While the staff was evaluating the Part 21 report, the licensee from a loss of

initiated several actions to improve the capability to recover to operate

SFP cooling. These actions included the following: (1) committing increase the

with the two SFPs cross-connected through the cask pit to committing to conduct

redundancy of cooling systems for the combined SFPs; (2)

reliability of the

testing and analyses that support assumptions regarding theanalyses that

SFP cooling assist mode of the RHR system; (3) completing installation of

support modifications and procedural changes; (4) completing conditions; and

instrumentation to improve the capability to monitor SFP that improve the

(5) completing changes to off-normal and emergency procedures

reliability of recovery from a loss-of-SFP-cooling event.

assessment

The staff used both deterministic and probabilistic safety a loss of

safety implications of events involving

techniques to evaluate the evaluation of the

SFP cooling. Because the staff did not consider a detailed of risk, the

effects of SFP boiling necessary, based on an initial assessment base

SFP boiling and

staff elected to quantitatively estimate the frequency of

decisions regarding further evaluations on that estimate.

to recover

The staff's deterministic engineering evaluation of the capability features of SSES

from or mitigate a loss of SFP cooling identified important characteristics

for modeling in the probabilistic safety assessment. These and outage

included the following: (1) on the basis of licensee commitments after a loss of

management procedures, the time to the onset of pool boiling

IN 93-83, Supp. 1 August 24, 1995 maintain the

cooling will exceed 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />; (2) natural circulation flow will [30'F] with the

temperature difference between the two pools less than 17C allowing a single

pools cross-connected through the common cask pit, thereby pool to

fuel pool cooling system of adequate capacity aligned to either errors, which

prevent boiling in both pools; (3) equipment failures and human

are explicitly modeled in the safety assessment, are the cooling significant failure

assist mode

modes for the normal SFP cooling system; and (4) the SFP or both pools

of the RHR system will provide a reliable means of cooling one

alignment is available.

when access to the reactor building for manual system

of reaching a

The safety assessment quantitatively estimated the frequency water vapor to

near-boiling condition, which could add significant heat and

The

the reactor building atmosphere, on the basis of the above information. to improve the

staff estimated that the actions the licensee has implemented

have reduced the near-boiling

capability to recover from a loss of SFP cooling

frequency from 7.OE-5 per plant-year to 2.OE-5 per plant-year.

LOOP,

The dominant sequences for near-boiling frequency involve an extended The

but sequences involving a LOCA or a shorter LOOP are also significant.

the reliance of the normal

dominance of sequences involving a LOOP reflects the limited

SFP cooling system on offsite sources of electrical power and of the RHR

availability of the RHR system for fuel pool cooling because

Sequences

system's primary reactor vessel decay heat removal function. because the RHR

involving a LOCA were identified as significant specifically

pool

system in the affected unit is assumed to be unavailable for fuel

cooling.

each unit at

Despite the relatively small fraction of an operating cycle that occurring during

SSES was assumed to be in a refueling outage,- the sequences

refueling outage periods that were examined dominated the near-boiling

shorter

frequency. Two factors contributed to this result: the relatively the practice of

time to reach boiling after a loss of SFP cooling because of

and the practice of removing systems

conducting full-core off-loads at SSES removal

associated with the outage unit that contribute to SFP decay heat

capability from service for maintenance during refueling outages.

and separate

To address generic concerns identified through the Part 21 report a special

concerns related to spent fuel storage pools identified during

Bulletin 94-01, inspection at a permanently shutdown reactor facility (see NRC at Dresden

"Potential Fuel Pool Draindown Caused by Inadequate Practicesbegun implementing

Unit 1," dated April 14, 1994), the staff has developed and actions:

a generic action plan. The generic plan includes the following storage pool

(1) a search and analysis of information regarding spent fuel

operation and design of spent fuel storage

issues, (2) an assessment of the the assessment

pools at selected reactor facilities, (3) an evaluation of of an

findings for safety concerns, and (4) selection and execution

based on the safety significance of the findings.

appropriate course of action that were

During these assessments, the staff will examine those features risk from loss- identified at SSES as important to the acceptably low level of

of-SFP-cooling events.

K>~/ K.) IN 93-83, Supp. 1 August 24, 1995 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.

IN 93-83, Supp. 1 August 24, 1995 written response.

This information notice requires no specific action or this notice, please

If you have any questions regarding the information inthe appropriate Office

contact one of the technical contacts listed below or

of Nuclear Reactor Regulation (NRR) project manager.

Dennis M. Crutchfield, ector

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 Attachment:

List of Recently Issued NRC Information Notices

A+1 4 ~# f?/.eV( QJckt7&

"'A'tachment

IN 93-83, Supp. 1 August 24, 1995 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

95-33 Switchgear Fire and 08/23/95 All holders of OLs or CPs

Partial Loss of Offsite for nuclear power reactors.

Power at Waterford

Generating Station, Unit 3

95-10, Potential for Loss of 08/11/95 All holders of OLs or CPs

Supp. 2 Automatic Engineered for nuclear power reactors.

Safety Features Actuation

95-32 Thermo-Lag 330-1 Flame 08/10/95 All holders of OLs or CPs

Spread Test Results for nuclear power reactors.

95-31 Motor-Operated Valve 08/09/95 All holders of OLs or CPs

Failure Caused by Stem for nuclear power reactors.

Protector Pipe Inter- ference

95-30 Susceptibility of Low- 08/03/95 All holders of OLs or CPs

Pressure Coolant Injection for nuclear power reactors.

and Core Spray Injection

Valves to Pressure Locking

94-66, Overspeed of Turbine- 06/16/95 All holders of OLs or CPs

Supp. 1 Driven Pumps Caused by for nuclear power reactors.

Binding in Stems of

Governor Valves

95-29 Oversight of Design and 06/07/95 All holders of OLs or CPs

Fabrication Activities for nuclear power reactors.

for Metal Components Used

in Spent Fuel Dry Storage

Systems

95-28 Emplacement of Support 06/05/95 All holders of OLs or CPs

Pads for Spent Fuel Dry for nuclear power reactors.

Storage Installations at

Reactor Sites

OL = Operating License

CP = Construction Permit

~ i k)

IN 93-83, Supp. 1 August 24, 1995 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.

orig /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 Attachment:

List of Recently Issued NRC Information Notices

DOCUMENT NAME: 9383SP1.IN

See nrevious concurrence

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NAME SJones* DSkeen* JShea* RSanders* GHubbard*

DATE 03/03/95 02/14/95 02/20/95 02/09/95 03/02/95 l

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NAME CMcCracken* GHolahan* RDennig* RKiessel* AChaffee*

.DATE 03/03/95 04/27/95 06/23/95 07/27/95 07/27/95 I l A

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OFFICE D:DR,1 NAME X DM ld

DATE 08/' /95 OFFICIAL RECORD 'COPY

IN 93-83, Supp. I

August xx, 1995 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.

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 Attachment:

List of Recently Issued NRC Information Notices

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DATE 103/03/95 104/27/95 106/23/95 107/27/95 07/27/95 __J

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NAME DMCrutchfield

4A1 DATE 08/ /95

OFFICIAL RECORD COPY

IN 93-83, Supp. 1 August xx, 1995 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.

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 Attachment:

List of Recently Issued NRC Information Notices

DOCUMENT NAME G:\DLS\iN95-KX--SFP X

See Drevious concurrence

OFFICE = SPLB Il OECB I DRPE l TECH ED I SC:SPL l l

NAME SJones* DSkeen* JShea* RSanders* GHubbard*

DATE 03/03/95 02/14/95 102/20/95 j02/09/95 03/02/95 OFFICE C:SPLB D:DSSA SCOECB I l ZECB

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DATE 07/ /95

OFFICIAL RECORD COPY

IN 93-83, Supp. I

June xx, 1995 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.

Brian K. Grimes, Director

Division of Project Support

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 Attachment:

List of Recently Issued NRC Information Notices

DOCUMENT NAME G:\DLS\IN95-XX.SFP

See nrevious concurrence

OFFICE SPLB lIOECB l DRPE Z TECH ED l SC:SPLB l

NAME SJones* DSkeen* JShea* RSanders* GHubbard*

DATE 03/03/95 02/14/95 02/20/95 02/09/95 03/02/95 OFFICE

NAME

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06/ /95 DATE 03/03/95 OFFICE D:DOPS I

NAME BKGrimes

DATE 06/ /95

OFFICIAL RECORD COPY

IN 93-83, Supp. 1 March xx, 1995 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.

Brian K. Grimes, Director

Division of Project Support

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 Attachment:

List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\DLS\IN95-XX.SFP

See Drevious concurrence

OFFICE SPLB OECB lDRPE TECH ED SC:SPLB

NAME SJones* DSkeen* JShea* RSanders* GHubbard*

DATE 03/03/95 02/14/95 02/20/95 02/09/95 03/02/95 OFFICE C:SPLB D:DSSA SC:OECB OECB C:OECB

NAME CMcCracken* GHolahan RDennig RKiessel AChaffee

DATE 03/03/95 102/ /95 02/ /95 02/ /95 OFFICE D:DOPS I

NAME BKGrimes

DATE 02/ /95

OFFICIAL RECORD COPY

K-, K)

IN 93-83, Supp. 1 February xx, 1995 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.

Brian K. Grimes, Director

Division of Project Support

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 Attachment:

List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\DLS\IN95-XX.SFP

OFFICE SPLB OECB DR4 TECH ED SC:SPLB

RSandersta GHubbaid/ 7 NAME SJones ' DSkeerfS J S

DATE 07/ 2/O5 02//q/95 02 /95 02/a1/95 0-

O/ /95 OFFICE C:ISPL) SC:OE OECB I lIC:OECI

ED:DSSA l

NAME CMkra e nL GHolahan RDennig RKiessel AChaffee

DATE  ?//95 02/ /95 02/ /95 02/ /95 02/ /95 OFFICE D:DOPS I

NAME BKGrimes

DATE 02/ /95 OFFICIAL RECORD COPY