Information Notice 1995-57, Risk Impact Study Regarding Maintenance During Low-Power Operation & Shutdown

Risk Impact Study Regarding Maintenance During Low-Power Operation & Shutdown
	ML031060179
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:	12/13/1995
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	;
References
	IN-95-057, NUDOCS 9512070252
	Download: ML031060179 (10)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 December 13, 1995 NRC INFORMATION NOTICE 95-57: RISK IMPACT STUDY REGARDING MAINTENANCE

DURING LOW-POWER OPERATION AND SHUTDOWN

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 inform licensees of an NRC Office of Research study on the effect of

system or equipment maintenance on boiling water reactor low-power and

shutdown core damage frequency. It is expected that licensees will review the

information for applicability to their facilities and consider actions, as

appropriate, to avoid risk-significant situations. However, suggestions

contained in this information notice are not NRC requirements; therefore, no

specific action or written response is required.

Discussion

The details of the study were published in October of 1994 in NUREG/CR-6166,

"Risk Impact of BWR Technical Specifications Requirements During Shutdown,"

(Accession No. 9411040066) which evaluated the risk impact of limiting

conditions for operation at low power and shutdown in the current technical

specifications for Grand Gulf Nuclear Station. These limiting conditions for

operation include allowed outage times, that is, the time allotted for system

maintenance outages.

A probabilistic model was developed for each of the eight plant operational

states, which are a reclassification of the five modes of operation defined in

the technical specifications. These eight plant operational states address

the range of plant conditions from refueling with the vessel head off and the

upper pool filled to full-power operation. The models were developed in

sufficient detail to allow for comparisons among the plant operational states.

Figure-of-merit measures included increase in conditional core damage

frequency. The measures explicitly accounted for the impact of test and

maintenance activities (e.g., allowed outage times and surveillance time

intervals) for both power and shutdown conditions.

NUREG/CR-6166 indicates that the increase in conditional core damage frequency

for taking a single train of standby service water out of service during

different plant operational states in an operating cycle can be substantial.

Single-train standby service water maintenance outages have been shown to

result in an increase in conditional core damage frequency during low power

9512070252 o ,r- A s

ASa

IN 95-57 December 13, 1995 1, 2, and 3)

and the first few days of hot shutdown (plant operational states frequency at

that is comparable to the increase in conditional core damage

full-power operation (plant operational state 0). During plant operational

and

state 4 (hot shutdown with shutdown cooling operating down to 0 psig) OF), the

plant operational state 5 (cold shutdown with temperatures < 200

core damage frequency could

study indicates that the increase in conditional on the

exceed that at full-power operation. NUREG/CR-6166 states, "Based

train

Increase in Conditional CDF (core damage frequency), single and bemultiplescheduled

maintenance on SSW (standby service water} should probably not

shutdown, for the late phases of hot shutdown and the early phases of coldalternative for

when the Alternate Decay Heat Removal System (ADHRS) is not an

decay heat removal."

If

This information notice requires no specific action or written response. contact

you have any questions about the information in this notice, please of

one of the technical contacts listed below or the appropriate Office

Nuclear Reactor Regulation (NRR) project manager.

Dennis M. Crutc ld rector

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Samuel S. Lee, NRR

(301) 415-1061 Arthur J. Buslik, RES

(301) 415-6184 John H. Flack, NRR

(301) 415-1094 Neal K. Hunemuller, NRR

(301) 415-1152 Attachment:

List of Recently Issued NRC Information Notices

A XIp g4yo k AJe

Attachment

IN 95-57 December 13, 1995 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

95-56 Shielding Deficiency in 12/11/95 All holders of OLs or CPs

Spent Fuel Transfer Canal for nuclear power reactors.

at a Boiling-Water Reactor

95-55 Handling Uncontained 12/06/95 All Uranium Recovery

Yellowcake Outside of a Licensees.

Facility Processing

Circuit

95-54 Decay Heat Management 12/01/95 All holders of OLs or CPs

Practices during for nuclear power reactors.

Refueling Outages

95-53 Failures of Main Steam 12/01/95 -All holders of OLs or CPs

Isolation Valves as a for nuclear power reactors.

Result of Sticking

Solenoid Pilot Valves

95-47, Unexpected Opening of a 11/30/95 All holders of OLs or CPs

Rev. 1 Safety/Relief Valve and for nuclear power reactors.

Complications Involving

Suppression Pool Cooling

Strainer Blockage

94-13, Control and Oversight of 11/28/95 All holders of OLs or CPs

Supp. 2 Contractors during Re- for nuclear power reactors.

fueling Activities and

Clarificaiton of Applica- bility of Section 50.120 of

Title 10 of The Code of

Federal Regulations to

Contractor Personnel

95-13, Potential for Data 11/22/95 All holders of OLs or CPs

Supp. 1 Collection Equipment to for nuclear power reactors.

Affect Protection System

Performance

OL - Operating License

CP = Construction Permit

IN 95-57 December 13, 1995 and the first few days of hot shutdown (plant operational states 1, 2, and 3)

that is comparable to the increase in conditional core damage frequency at

full-power operation (plant operational state 0). During plant operational

state 4 (hot shutdown with shutdown cooling operating down to 0 psig) and

plant operational state 5 (cold shutdown with temperatures < 200 'F), the

study indicates that the increase in conditional core damage frequency could

exceed that at full-power operation. NUREG/CR-6166 states, *Based on the

Increase in Conditional CDF {core damage frequency), single and multiple train

maintenance on SSW {standby service water) should probably not be scheduled

for the late phases of hot shutdown and the early phases of cold shutdown, when the Alternate Decay Heat Removal System (ADHRS) is not an alternative for

decay heat removal."

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 project manager.

orig /s/'d by DMCrutchfield

Dennis M. Crutchfield, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical Contacts: Samuel S. Lee, NRR

(301) 415-1061 Arthur J. Buslik, RES

(301) 415-6184 John H. Flack, NRR

(301) 415-1094 Neal K. Hunemuller, NRR

(301) 415-1152 Attachment: List of Recently Issued NRC Information Notices

See previous concurrences.

TechEd reviewed this document 9/27/95.

DOCUMENT NAME: 95-57.IN

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OFFICE Tech Contacts* D/DSSA PRAB/DST:RES E C/PECB/DRPM

NAME GHolahan* MCunningham* AChaffee*

DATE 10/04/95 11/08/95 11/13/95

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DATE 112/11 /95 OFFICIAL RECORD COPY

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IN 95-XX

October xx, 1995 cold shutdown or when the alternate decay heat removal system is not capable

of serving as a backup to the normal decay heat removal system.

It is expected that licensees will review the information in NUREG/CR-6166 for

applicability to their facilities and consider actions, as appropriate, to

avoid risk-significant situations. However, suggestions contained in this

information notice do not constitute NRC requirements; therefore, no specific

action or written response is required. 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.

Dennis M. Crutchfield, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical Contacts: Samuel S. Lee, NRR

(301) 415-1061 Arthur J. Buslik, RES

(301) 415-6184 John H. Flack, NRR

(301) 415-1094 Neal K. Hunemuller, NRR

(301) 415-1152 Attachment: List of Recently Issued NRC Information Notices

See previous concurrences.

DOCUMENT NAME: G:\NKH\INFONOTI.BST

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DATE 9/25/95 9/27/95 9/28/95 9/29/95 10/4/95 OFFICE PRAB:DST:REEOECB:DOPS

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IN 95-XX

December xx, 1995 that is comparable to the increase in conditional core damage frequency at

full-power operation (plant operational state 0). During plant operational

state 4 (hot shutdown with shutdown cooling operating down to 0 psig) and

plant operational state 5 (cold shutdown with temperatures < 200 OF), the

study indicates that the increase in conditional core damage frequency could

exceed that at full-power operation. NUREG/CR-6166 states, *Based on the

Increase in Conditional CDF (core damage frequency), single and multiple train

maintenance on SSW {standby service water) should probably not be scheduled

for the late phases of hot shutdown and the early phases of cold shutdown, when the Alternate Decay Heat Removal System (ADHRS) is not an alternative for

decay heat removal."

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.

Dennis M. Crutchfield, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical Contacts: Samuel S. Lee, NRR

(301) 415-1061 Arthur J. Buslik, RES

(301) 415-6184 John H. Flack, NRR

(301) 415-1094 Neal K. Hunemuller, NRR

(301) 415-1152 Attachment: List of Recently Issued NRC Information Notices

See previous concurrences.

DOCUMENT NAME: G:\NKH\INFONOTI.BST _o

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OFFICE SPSB:DSSA IETech. Ed SPSB:DSSA E BE

NAME SLee;cf* MMejac* JFlack* EButcher* GHolahan*

DATE 9/25/95 9/27/95 9/28/95 l 9/29/95 10/4/95

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DATE 11/08/95 111/13/95 11113 9---5 I. / /95.

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IN 95-XX

December xx, 1995 that is comparable to the increase in conditional core damage frequency at

full-power operation (plant operational state 0). During plant operational

state 4 (hot shutdown with shutdown cooling operating down to 0 psig) and

plant operational state 5 (cold shutdown with temperatures

200 OF), the

study indicates that the increase in conditional core damage frequency could

exceed that at full-power operation. NUREG/CR-6166 states, "Based on the

Increase in Conditional CDF [core damage frequency}, single and multiple train

maintenance on SSW {standby service water) should probably not be scheduled

for the late phases of hot shutdown and the early phases of cold shutdown, when the Alternate Decay Heat Removal System (ADHRS) is not an alternative for

decay heat removal."

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.

Dennis M. Crutchfield, Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical Contacts: Samuel S. Lee, NRR

(301) 415-1061 Arthur J. Buslik, RES

(301) 415-6184 John H. Flack, NRR

(301) 415-1094 Neal K. Hunemuller, NRR

(301) 415-1152 Attachment: List of Recently Issued NRC Information Notices

See previous concurrences.

DOCUMENT NAME: G:\NKH\INFONOTI.BST

To receive a copy of this document, Indicate In the box: 'C - Copy without attachment/enclosure E - Copy with attachment/enclosure N' = No copy

OFFICE SPSB:DSSA E Tech. Ed SPSB:DSSA I E SPSB:DSSA I D:DSSA IE

NAME SLee;cf* MMejac* JFlack* EButcher* IGHolahan*

DATE 9/25/95 9/27/95 9/28/95 9/29/95 10/4/95 OFFICE IPRAB:DST:RES IE IPECB:DRPM IE IPECB:DRPM IE IDRPM:NRR IE l

NAME JMCunningham* lNHunemuller* IAChaffee* IDCrutchfield

DATE 111/08/95 111/13/95 111/13/95 / /95 OFFICIAL RECORD COPY

KY

IN 95-XX

October xx, 1995 when the alternate decay heat removal system (e.g., Grand Gulf) is not capable