ML20247F033
| ML20247F033 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 05/18/1989 |
| From: | Morris K OMAHA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LIC-89-284, NUDOCS 8905260401 | |
| Download: ML20247F033 (3) | |
Text
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Omaha Public Power District 1623 Harney Omaha. Nebraska 68102 2247 402/536-4000 May 18, 1989 LIC-89-284 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station PI-137 Washington, D.C.
20555
References:
1.
Docket No. 50-285 2.
Letter from NRC (J. A. Calvo) to OPPD (R. L. Andrews) dated November 17, 1987 3.
Letter from LeBoeuf, Lamb, Leiby & MacRae to NRC (H. R.
Denton) dated July 17, 1986 Gentlemen:
SUBJECT:
Response to Request for Further Information Regarding Reactor Vessel Embrittlement for Further Consideration of Application for Life Extension Omaha Public Power District (OPPD) was requested in Reference 2 to provide additional information on reactor vessel embrittlement to support a previous request for a five year license extension (Reference 3).
If 10 CFR50.61 (The PTS Rule) is amended to include the radiation embrittlement equation of Regulatory Guide 1.99, Revision 2, the Commission noted that Fort Calhoun is expected to exceed the Pressurized Thermal Shock (PTS) screening criteria before the end of license life (June 7, 2008) and, therefore, advised OPPD to consider a neutron flux reduction program.
To this end, OPPD is continuing studies in many areas of vessel embrittlement including:
Extreme low radial leakage fuel management study Regulatory Guide 1.154 analysis Reactor Vessel Annealing OPPD commissioned a study to evaluate the impact of fluence reduction for the Extreme Low Radial Leakage (ELRL) fuel management strategy employed for Cycle
- 10. Combustion Engineering (CE) was contracted during November, 1988 to provide a detailed fluence distribution evaluation for Cycle 10 using the DOT
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V. S. Nuclear Regulatory Commission CIC-89-284 Page 2-4.3 code. Results of the CE analysis were presented to OPPD on February 6, 1989.
Subsequent calculations based upon the Cycle 10 DOT 4.3 fluence analysis revealed the following:
1.
If ELRL fuel management strategies equivalent to Cycle 10 are used for all cycles starting with Cycle 13, Fort Calhoun Station could operate until the end-of-license life.
2.
In order to reach end-of-license life plus a five-year extension, it would be necessary to initiate a fuel management strategy with greater f7ux reduction than Cycle 10.
3.
Delaying the ELRL fuel management from Cycle 13 to Cycle 14 does not significantly impact the amount of increased fluence reduction greater than the reduction necessary beginning in Cycle 13.
Based on the knowledge that has been gained from research, industry experience and past experience, implementation of an extreme low radial leakage fuel management strategy is planned for Cycle 14.
Fort Calhoun Station is currently in Cycle 12.
In addition, OPPD is currently participating in Phase 1 of a CE Owners Group task to perform a Regulatory Guide 1.154 analysis for Fort Calhoun.
Regulatory Guide 1.154 (Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors) describes the recommended methods to be used in performing safety analyses which are required by the rule for any plant that wishes to operate beyond the PTS screening criteria. This task is divided into three phases and is scheduled for completion in 1992.
The first phase is to finalize a plan to perform the analysis and present the plan to the NRC for preliminary approval of the analysis methods. The second phase of the task will be to perform an analysis of a generic CE plant using the approved methodology. The third and final phase of the owners group task would perform a plant specific analyses for the i
plants participating in the task and present the results of the analyses to the NRC.
If it is decided not to continue with the CE Owners Group effort, a PTS scoping risk assessment would be employed to evaluate the risk to Fort Calhoun
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Station due to PTS events and show what modifications to plant equipment, i
systems, or procedures would be necessary for Fort Calhoun Station to be i
operated above the screening criteria. This assessment would also indicate whether or not a full Regulatory Guide 1.154 analysis would be successful if l
one was performed.
Finally, the industry's progress for the development of vessel annealing capabilities is being monitored so that, if vessel annealing becomes necessary, i
l OPPD will have the most current technology available. OPPD is also participating in industry programs such as the EPRI Reactor Vessel Embrittlement Management Program to gain knowledge of inoustry progress and new research regarding the management of vessel embrittlement.
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.U. S. Nuclear Regulatory' Commission
. -- *L1C-89-284 Page 3 As the above studies are concluded, OPPD will submit the results to the staff to support further review of the OPPD five year extension of license applica-tion.
If you have further questions on the matter, please contact me or members of my staff.
Sincerely,
[
orris Division Manager Nuclear Operations KJM/jak c:
LeBoeuf, Lamb, Leiby & MacRae R. D. Martin, NRC Regional Administrator P. D. Milano, NRC Project Manager P. H. Harrell, NRC Senior Resident Inspector d
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