Information Notice 1987-65, Plant Operation Beyond Analyzed Conditions

Plant Operation Beyond Analyzed Conditions

ML031130392
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, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	12/23/1987
From:	Rossi C Office of Nuclear Reactor Regulation
To:
References
IN-87-065, NUDOCS 8712170119
Download: ML031130392 (7)

IN 87-65 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON D.C. 20555 December 23, 1987 NRC INFORMATION NOTICE NO. 87-65: PLANT OPERATION BEYOND ANALYZED

CONDITIONS

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

This information notice is being provided to alert addressees to potential

problems resulting from operating a plant beyond its analyzed basis. The

safety concerns of the particular circumstances described in this information

notice are high temperature inside containment and insufficient post-LOCA

cooling of safety systems. It is expected that recipients will review the

information for applicability to their facilities and consider actions, as

appropriate, to avoid similar problems. However, suggestions contained in

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

specific action or written response is required.

Description of Circumstances

Arkansas 1 (ANO-1). During normal operation on August 7, 1987, it was found

that the containment temperatures were significantly higher than the tempera- tures assumed in the accident analyses in the final safety analysis report

(FSAR, including updates) and equipment qualification program. In the FSAR,

a design temperature of 110OF was assumed for safety analysis of containment

integrity and 1201F was assumed for equipment qualification during normal

103OF to 1650F with one local

service life. Measured temperatures ranged from

"hot spot" of 1831F about the "A" steam generator. The licensee had observed

such temperatures since plant startup in 1974.

Crystal River 3. During an inspection, it was found that the temperature of 0

the ultimate heat sink (UHS), the Gulf of Mexico, was above the value of 85 F

assumed in the FSAR analysis for heat removal capability after a loss-of- coolant accident (LOCA). The Technical Specifications (TS) permit a UHS

temperature of 1050F. The plant has been 0operating within the TS limit but

beyond the design-basis temperature of 85 F assumed in the accident analysis.

AZ 8712170119

IN 87-65 December 23, 1987 Discussion: l, Because'ANO-1 had been operating at elevated containment temperatures for

extended periods, the NRC staff had several concerns:

1. The plant had been operating beyond its analyzed basis with regard to

post-accident (LOCA) containment performance because the initial condi- tions assumed in the analysis were exceeded.

2. The higher temperature implies accelerated aging of equipment required for

post-accident safe shutdown in accordance with regulation 10 CFR 50.49 on

equipment qualification.

3. The higher temperature may cause deterioration of the concrete structure.

In response to the NRC staff concerns, the licen'see submitted an analysis of

the safety implications of the elevated containment temperatures and identified

both near term and long term actions to justify continued operation.

In general, the FSAR contains design bases, operational limits, and analyses of

structures, systems, and components for ensuring the safety of the facility.

It is a statement-by the applicant/licensee of how it intends to comply with

NRC requirements. This statement is reviewed by the NRC to form the bases for

the operating license. The analysis of containment performance following a

- design-hasis accidet (for-example, a4LOCA) depands-on-certal-n-assumed-Thit-al---

conditions. Exceeding these conditions may invalidate the analysis and thereby

raise concerns regarding the maintenance of containment integrity following an

accident.

In accordance with the "100 C rule," which may be used to calculate qualified

life, an increase of 100 C (180 F) over the initially assumed temperature reduces

the qualified life by 50 percent. Under these circumstances, equipment that is

relied on in the event of a design-basis accident may not reliably perform

its safety function when required.

In the case of Crystal River 3, the concern was consistency between the FSAR

and the TS. Regulation 10 CFR 50.36 requires that the TS be derived from the

analyses in the safety analysis report. Since the plant has been operating

beyond the assumed design-basis temperature for the UHS, the adequate transfer

of post-accident heat loads from safety-related structures, systems, and

components was in question.

Attachment

IN 67-65 December 23, 1987 face 1 of I

LIST OFRECENTLY ISSUED

NRCINFORWATION NOTICES 1387 I--_s . ..

Notice No. Sublect Isuenev Issued to

87464 Conviction for Falsatication 22/22/67 All nuclear power

of Security Training Redords reactor facilities

holding an OLor CP

and ll major fuel

facility licensews

67-35, Supp. 1 Reactor Trip Breaker 12/26/87 All holders of OLs

Westinghouse Nodal OS-416, or Cps for nuclear

- Failed to Open OnManual ,ower reactors.

Initiation From the Control

Room

87-63 Inadequate Net Positive 11/9/87 All holders of OLe

Suction Hoed in Low Pressure or CPs for nuclear

Safety Systems por reactors.

87-62 1echanical Failure of 12/6/67 All holders of OLt

Indicating-Type Fuses or CP for nuclear

power reactors.

87-61 Failure of Westinghouse 1l167 All holders of OLs

W-2-Type Circuit Breaker or CPs for nuclear

Coll Switches. power reactors.

87-60 DCpressurization of Reactor 12/4/67 All holders of OLt

Coolant systems In or CPs for PWRs.

Pressuri od-Water Reactors

65-lo0. Degradation of Reactor l1/97 All holders of OLs

Supp. Cool2nt Syst Pressure or CPs for nuclear

0oundary Iesulting from power reactors.

6oric Acid Corrosion

87-59 Potential RHRPunp Loss 21117/87 All holders of OLS

or cps for nuclear

power reactors.

OL : Operating License

CP* Construction Permit

UNITED STATES IPOSTAGE 6 FEES PAOID

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555 OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

IN 87-65 December 23, 1987 No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the technical

contact below or the Regional Administrator of the appropriate regional office.

ar es E. Rossi, Director

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical Contacts: C. Li, NRR

(301) 492-9414 Vern Hodge, NRR

(301) 492-8196

. Attachment: List of Recently Issued NRC Information Notices

MI

o 1 '!! i tLS c..3 U1 11I,;- . . _ ~-:: i : . . .<._I:,\_

,: -: I _n~.-

.:

IN 87-65 December 23, 1987 No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the technical

contact below or the Regional Administrator of the appropriate regional office.

Charles E. Rossi, Director

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical Contacts: C. Li, NRR

(301) 492-9414 Vern Hodge, NRR

(301) 492-8196 Attachment: List of Recently Issued NRC Information Notices

I

The draft of this information notice was transmitted to DOEA by DEST in a

memorandum from L. Shao dated 10/26/87.

• SEE PREVIOUS CONCURRENCES

• OGCB:DOEA:NRR 'PPMB:ARM *SAD/DEST:NRR ,C/OGCB: DOEA: NRR

CVHodge TechEd AThadani CHBerlinger

11/27/87 12/11/87 12/13/87 12/16/87

IN 87-XX

December xx, 1987 No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the technical

contact below or the Regional Administrator of the appropriate regional office.

Charles E. Rossi, Director

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical Contacts: C. Li, NRR

(301) 492-9414 Vern Hodge, NRR

(301) 492-8196 Attachment: List of Recently Issued NRC Information Notices

The draft of this information notice was transmitted to DOEA by DEST in a

memorandum from L. Shao dated 10/26/87.

• SEE PREVIOUS CONCURRENCES s i0

• OGCB:DOEA:NRR PPMB:ARM SAD/DEST:NRR C/OGCB:DOEi k:NRR D/DOEA:NRR

CVHodge TechEd6 AThadani CHBerlingei r CERossi

11/27/87 121/i /87 12/1 /87 12//a/87 12/ /87

IN 87-XX

November xx, 1987 No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the technical

contact below or the Regional Administrator of the appropriate regional office.

Charles E. Rossi, Director

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical Contacts: C. Li, NRR

(301) 492-9414 Vern Hodge, NRR

(301) 492-8196 Attachment: List of Recently Issued NRC Information Notices

a

OGCB:DOEA:NRR PPMB:ARM C/OGCB: DOEA:NRR D/DOEA:NRR

CVHodge TechEd CHBerlinger CERossi

11/j1/87 11/ /87 11/ /87 11/ /87