Information Notice No. 87-65: Plant Operation Beyond Analyzed Conditions

                                                              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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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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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