Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
IN 87-65
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON D.C. 20555
December 23, 1987
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 110 F was assumed for safety analysis of containment
integrity and 120 F was assumed for equipment qualification during normal
service life. Measured temperatures ranged from 103 F to 165 F with one local
"hot spot" of 183 F 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
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 105 F. The plant has been operating within the TS limit but
beyond the design-basis temperature of 85 F assumed in the accident analysis.
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December 23, 1987
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Discussion:
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 licensee 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-basis accident (for example, a LOCA) depends on certain
assumed initial 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 "10 degree C rule," which may be used to calculate
qualified life, an increase of 10 degrees C (18 degrees 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.
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December 23, 1987
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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