ML20136D600
| ML20136D600 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/06/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20136D595 | List: |
| References | |
| NUDOCS 9703120413 | |
| Download: ML20136D600 (3) | |
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't UNITED STATES
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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 001
- ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION j
RELATED TO AMENDMENT NO.188 TO FACILITY OPERATING LICENSE NO. DPR-16 GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION LOCKET N0. 50-219
1.0 INTRODUCTION
By letter dated August 23, 1996, General Public Utilities (GPU) Nuclear Corporation (the licensee) submitted a request for changes to the Oyster Creek Nuclear Generating Station (OCNGS) Technical Specifications (TS).
GPU proposed to revise TS 3.3.A (i), (ii), (iii), and (iv); the pressure-l temperature (P-T) limit curves in Figure 3.3.1; TS 4.3.A; and the Bases and references for the aforementioned sections.
GPU supplemented this submittal by letter dated January 8,1997, in response to a request for additional information (RAI) from the NRC staff of November 27, 1996.
The January 8, 1997, letter provided clarifying information within the scope of the original application and did not change the staff's initial proposed no significant hazards consideration determination.
The staff has reviewed the information provided by the licensee and has determined that the licensee used methodologies consistent with the requirements of Title 10 of the Code of Federal Regulations Part 50 (10 CFR Part 50), Appendix G, Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Code, and Standard Review Plan (SRP) Section 5.3.2, " Pressure-Temperature Limits."
The requirements and guidelines that have been established or endorsed by the NRC concerning pressure versus temperature limitations on the reactor vessel environment provide adequate margins of safety against brittle fracture of the reactor vessel during all modes of operation. Appendix G of 10 CFR Part 50 identifies three specific vessel operating conditions to consider when developing P-T limit curves: vessel hydrostatic or leak-rate testing; normal operation with the core not critical, and normal operation with the core critical.
The licensee's submittal addressed each of these operating conditions and proposed a P-T limit curve for each.
All regions of the reactor vr',sel must be evaluated to ensure vessel integrity during each of the operating conditions noted above. As addressed in 10 CFR Part 50, Appendix G, two regions of the reactor vessel are specifically 9703120413 970306 PDR ADOCK 05000219 P
. identified relating to the establishment of these limitations:
the closure flange region and the vessel beltline region.
In addition, Appendix G to Section XI of the ASME Code recommends methodologies for the analysis of geometrically complex vessel regions, namely the vessel lower head and the feedwater nozzle region. The analysis submitted by the licensee considered each of these regions. The licensee proposed that the P-T limit curve for each operational condition then be constructed as a composite of the most limiting values for all pressures of interest. This resulted in the P-T limit curves for each operational condition being controlled at low pressures by consideration of the reactor vessel flange region and at high pressures by consideration of the reactor vessel beltline. The reactor ves:e1 bottom head was the controlling region for vessel hydrostatic / leak rate testing at intermediate pressures. The feedwater nozzle region controlled both of the l
normal operation curves in the intermediate pressure regime.
The regulatory requirements concerning the development of P-T Limits are.
J contained in Appendix G to 10 CFR Part 50 and provide the general basis for these limits.
The requirements of 10 CFR Part 50, Appendix G, specifically state that the P-T limits for a vessel must be at least as conservative as limits obtained by following the methods of analysis and the margins of safety of Appendix G to Section XI of the ASME Code. The staff has also established guidance for the development of P-T limits in SRP Section 5.3.2.
Staff guidance on the calculation of material embrittlement as a result of neutron radiation has been published in Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations."
j 2.0 EVALUATION The staff first reviewed the nil-ductility reference temperatures (RTut) proposed by the licensee for the reactor vessel non-beltline regions and determined that they were acceptable. Due to low neutron fluence levels these materials do not experience any irradiation-induced shift in RT through 32 effective full power years (EFPYs).
The staff then confirmed Nat the values the irradiation-of the adjusted RT induced shift in RI (a combination of the initial RTand a margin term) for the limiNn,g (Plate G-8-6) were N5,152, and 158 *F as calculated by the licensee at 22, 27, and 32 EFPYs, respectively. The shift in RT for the beltline materials was calculated by the staff in accordance with rig *ulatory Guide 1.99, Revision 2.
With this information, the staff was able to verify the licensee's P-T limit curves using the methodology of SRP Section 5.3.2 and Appendix G to Section XI of the ASME Code.
Initially the staff calculated the P-T limit curves for the limiting beltline material at 22, 27, and 32 EFPYs for vessel hydrostatic / leak rate testing and for the limiting normal operation transient, a 100 *F/hr cooldown. The staff then reviewed the licensee's calculations for the non-beltline lower head and feedwater nozzle regions from the RAI response and verified those results.
Finally, the staff constructed the composite P-T curves for each operational condition from the information noted above and the minimum temperature constraints based on the limiting RT, of the closure g
flange region as required by 10 CFR Part 50, Appendix G he m
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O e The staff determined that the P-T limits proposed for vessel hydrostatic / leak rate testing by the licensee were acceptable to 22, 27, and 32 EFPYs. The staff also confirmed that the P-T limits curves for 22, 27, and 32 EFPYs submitted by the licensee for normal operation under core critical and non-critical conditions were consistent with those generated using the methodologies of SRP Section 5.3.2 and Appendix G of the ASME Code.
In addition, normal boiling water reactor operations are constrained to saturated water conditions, and these conditions are very conservative with respect to the normal operation limits as calculated by the acceptable methodologies listed above.
The staf? has reviewed the information provided by the licensee in the documer.ts listed above.
The staff has determined that the licensee used methodologies consistent with the 10 CFR Part 50, Appendix G, Appendix G to Section XI of the ASME Code, Regulatory Guide 1.99, Revision 2, and SRP Section 5.3.2.
Therefore, these changes to the Oyster Creek Nuclear Generating Station TS are acceptable.
3.0 STATE CONSULTATION
In accordance with the Comission's regulations, the New Jersey State official was notified of the proposed issuance of the amendment.
The State official had no coments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a 2
proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (61 FR 47977). Accordingly, the amendment meets the eligibility criteria for
~categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prcpared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Comission's regulations, and (3) the issuance of the amendment will not be inimical to the comon defense and security or to the healtn and safety of the public.
Principal Contributor:
M. Mitchell i
Date: March 6, 1997