ML20091J287
| ML20091J287 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/27/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20091J284 | List: |
| References | |
| NUDOCS 9201100252 | |
| Download: ML20091J287 (5) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIO'l RELATED TO AMENDMENT NO. 77 TO FACILITY OPERATING LICENSE NO. NPF-43 DETROIT EDISON COMPANY FERM1-2 DOCKET NO. 50-341
1.0 INTRODUCTION
Sy letter dated February 21, 1991, the Detroit Edison Company (DECO or the licensee) requested amendment to the Technical Specifications (TS) appended to facility Operating License No. NPF-43 for fermi-2. TH prorosed amendment ch;nge provides pressure. D perature (P/T) limits for the n?ctor coolant system and the reactor pecure vessel for the Fermi-2 Techrdcal Specifications, Section 3.4.6.1.
The propused P/T limits were requcsted for 32 effective full power years (EFP"i and were developed using Regulatory Guide (RG) 1.99, Revision 2.
Ge
- Letter 88-11. "NRC Position on Radiation Embrittlement of Reactor Vessel Ma.1als and its impact on Plant Operations," recommends RG 1.99, Rev. 2, be used in solculating P/T limits, unless the use of different methods cai, he justified. The P/T limits provide for the operation of the reactor coolant systen during heatup, cooldown, criticality, and hydrotest.
To evaluate the P/T limits, tne staU uses the following NRC regulations and guidance:
Appendices G and H of 10 CFR Part 50; the ASTM Standerds and the ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2);
RG 1.99, Rev. 2; Standard Review Plan (SRP) Section 5.3.2; and Gene:fc Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.3fi to provide Technical Sp(ecifications for the operation of the plant.
In particular, 10 CFR 50.35 c)(2) requires that limiting conditions of operation be included in the Technical Specifications. The P/T limits are among the limiting conditions of operation in the Technical Specifications for all commercial-nuclear plants in the U.S.
Appendices G and H of 10 CFR Part 50 describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in setting P/T limits. An acceptable method for constructing the P/T limits h described in SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements i
for reactor vessel materials in accordanco with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested in accordance l
with Appendix H of 10 CFR Part 50. Ap p ndix H, in turn, refers to ASTM Standards.
l These tests define the extent of vesul en6rittlement at the time of capsule withdrawal in terms of the increase 4 reference temperature. Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and p ? H"DOLnV U =11 w t
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-2 Charpyuppershelfenergy(USE). Generic Letter 88-11 requested that licensees and permittees use the methods in RG 1.99, Rev. 2, to predict the effect of neutron irradiation on reactor vessel materials. This guide defines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting fron neutron irradiation, and a margin to account for uncertainties in the prediction method.
Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules frem the reactor vessel.
Appendix H refers to the ASTM Standards which, in turn, require that the capsules be installed in the vessel before startup and that they contain test specimens made from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Fermi-2 reactor vessel. The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 32 EFPY was the lower shell longitudinal welds 2-307A, B, and C with 0.26% copper (Cu), 0.87% nickel (N1), and an initial RT o -104.
ndt The licensee has not yet removed any surveillance capsule from Fermi-2 except that the flyx wire dosimeter attached to Capsule 3 was removed. There are three specimen capsules in the surveillance program and the first capsule will be ren.oved at 8 EFPY. The plant is at 2.5 EFPY in October 1991. All surveillance capsules contain Charpy impact spe::imens and tensile specimens made-from base metal, weld metal, and HAZ metal.
For the limiting beltline material, weld 2-307A, B, and C, the staff calculated l
theARTtobe65.2' Tat 1/4T(T=reactorvesselbeltlinethickness)gnd26.2*F for 3/4T at 32 The staff used a neutron fluence of 3.6E17 n/cm at 1/4T and1.7E17n/cm{FPY.
l at 3/4T. The ART was determined by Section 1 of RG 1.99, Rev.
2, because no surveillance capsules have been removed from the Fermi-2 pressure vessel.
The licensee used the method in RG 1.99 Rev. 2, to calculate an ART of 65.3*F at 32 EFPY at 1/4T for the same limiting weld metal.
The staff judges that the licensee's ART 65.3'F is more conservative than the staff's ART of 65.2"F, and l
it is acceptable. Substituting the ART of 65.2*F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cocidown, and hydro-test meet the beltline material requirements in Appendix G of 10 CFR Part 50.
In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposes P/T limits based on the reference temperature for the reactor vessel closure flange materials.Section IV.2 of Appendix G states that when the pressure exceeds 20% of the preservice system hydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed the reference temperature of the material in those regions by at least 120 F for normal operation and by 90'F for hydrostatic pressure tests and leak tests.
Paragraph IV.A.3 of Appendix G states "an exception may be made for boiling
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. water reactor vessels when water level is within the normal range for power operation and the pressure is less then 20 percent of the preservice system hydrostatic test pressure.
In this case the minimum permissible temperature is 60*F (33*C) above the reference temperature of the closure flange regions that are highly stressed by-the bolt preload." Based on the flange referance temperature of 11 F, the staff has determined that the proposed P/T limits, satisfy Section IV.2 of Appendix G.
Section IV B of Appendix G requires that the predicted Charpy USE at end of life be above 50 ft-lb.
The material with the lowest unitradiated USE is weld 2-307A, B, and C with 67 ft-lb.
The staff calculates that at 32 EFPY and at 1/4T, weld 2-307A, B, and C will have a USE of 50.4 ft-lb.
This is greater than 50 ft-lb and, therevore, is acceptable.
Table 4.4.6.1.3-1 has been modified to reflect the lead factor used for the updated P-T curves.
The lead factor better reflects the variation between the peak fluence location and capsule location.
The staff has reveiwed the change and finds the change is acceptable since the new lead factor specified in consistent with those previously found to be acceptable in vessel configurations similar to Fermi 2 (e.g., Susquehanna 1 and Peach Bottom 2).
The licensee also proposes to change the TS Bases section B 3/4.4.6, Pressure-Temperature Limits.
These changes are necessary to reflect the updated RG 1.99 Revision 2 methodology and the conclusions of this evaluation.
In addition, editorial changes ar2 made to correct references to the UFSAR and to eliminate detailed information typicelly located in the USFAR.
These Bases changes are acceptable since they are either cons. stent
-with the revised methodology or editorial in nature.
The staff concludes that the proposed P/f limits for the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 32 EFPY because the limits conform to tue requirements of Appendices G and H sf 10 CFR Part 50.
The proposed P/T limits also satisfy Generic Letter 88-11 because the method in RG 1.99, Rev. 2 was used to calculate the ART.
Therefore, the proposed P/T limits may be incorporated into the Fermi-2 Technical Specifications.
3.0 REFERENCES
1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials, Revision 2. May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2:
Pressure-Temperature
-Limits 3.
February 21, 1991, Letter from W. S. Orser (DE) to USNRC Document Control Desk,
Subject:
Proposed Technical Specification Change (License Amendment) -
Pressure / Temperature Limits a
4
4.0 STATE CONSULTATION
in accordance with the Commission's regulaticns, the Michigan State of ficial was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes in surveillance requirements.
The staff has determined that the amendment involves no significant increase in the amnunts, and nc significant change in the types, of any effluents which may be released offsite, and that there is no significant increase in individual or cumulative occupaticnal radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (56 FR 57694).
Accordingly, the amendment meets the eligibility criteria for categorical axclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmentai impact statement 01 environmental assessment need be prepared in connection g
with the issuance of the amendment.
6.0 CONCLU5 ION The staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the haalth 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defensa and security or to the health and safety of the public.
Principal Contributor:
J. Tsao Date: December 27, 1991
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4 DATED: December 27, 1991 AMENDMENT i;0. 77 TO FACILITY OPERATING LICENSE N0. NPF-43 FERMI-2 7 Docket File 1 NCC & Local PDRs PDIII-1 Reading Fermi Plant File B. Boger-J. Zwo'inski L. Marsh P. Shuttleworth J. Stang OGC-WF D. Hagan, 3302 MNBB G. Hill (4), P-137 Wanda Jones, MNCB-7103 C. Grimes, 11/F/23 J. Tsao 7/0/4 ACRS (10,'
GPA/PA OC/LFMB W. Shafer RIII