ML20058C422
| ML20058C422 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 10/19/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058C417 | List: |
| References | |
| NUDOCS 9011010222 | |
| Download: ML20058C422 (5) | |
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1 UNITED STATES
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NUCLEAR REGULATORY COMMISSION -
'A7 y WA sHING TON, D. C. 70655
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
36 TO FACILITY OPERATING LICENSE NPF-68 AND AMENDMENT NO. 16 TO FACILITY OPERATING LICENSE NPF t GEORGIA POWER COMPANY. ET AL.
DOCKETS NOS. 50-424 AND 50-425-V0GTLE ELECTRIC GENERATING PLANTc UNITS 1 AND 2
1.0 INTRODUCTION
By [[letter::ELV-00571, Application for Amends to Licenses NPF-68 & NPF-81,revising Tech Spec 3.4.9 & Associated Bases,To Decrease Heatup & Cooldown Curves,For Unit 1,from 16 EFPY to 13 Efpy,Based on Westinghouse Analysis Per Generic Ltr 88-11.W/o Encl|letter dated July 28, 1989]], Georgia Power Company,-et al. (the licensee),
requested a change to the Technical Specifications (TSs) for the Vogtle Electric Generating Plant (VEGP), Units 1 and 2. =The proposed change-responds to Generic Letter (GL) 88-11. "NRC Position on Radiation' Embrittlement of Reactor Vessel Materials and Its Effect on Plant 0perations" and revises TS 3.4.9, " Pressure / Temperature Limits'," and the associated bases concerning the number of effective full power years (EFPY) for which the Unit I heatup and coolclown curves are applicable. Specifically, TS Figures 3.4-2a and 3.4-3a,-
as well as pages B3/4 4-8 and 4-10, are revised to reflect the change of EFPY from 16 to'13. The Vogtle Unit 2 pressure / temperature'(P/T) limits for"16 EFPY remain unchanged.
1 To evaluate the P/T limits, the NRC staff used the.following NRC regulations and guidance: Appendices G and H of 10 CFR Part 50; the ASTM Standards 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 Generic Letter 88-11.
Each licensee authorized to operate a nuclear power reactor is required by-10 CFR 50.36 to provide TSs for the operation of the plant.- in particular, 10 CFR 50.36(c)(2) requires that limiting conditions of operation be-included I
in the TSs. The P/T limits are among the. limiting. conditions of operation in L
the TSs for all commercial nuclear plants in the U.S.
Appendices G cad 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 i
limits. An acceptable method for constructing the P/T limits-is described in-SRP Section 5.3.2.
Appendix G of 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME Code and, in particular, that the beltline materials in the surveillance capsules be tested in accordance with Appendix H of 10 CFR Part 50. Appendix H, in turn, refers to ASTM Standards. These tests define the extent of vessel embrittlement at-the time of capsule withdrawal in terms of the increase in reference temperature.
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Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (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 unitradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction method.
I Appendix H of 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from 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 NRC staff evaluated the effect of neutron irradiation embritt.lement on each l
beltline material in the Vogtle 1 reactor vessel.
The amount of irradiation l
embrittlement was calculated in accordance with Section 1 of Regulatory Guide (RG) 1.99, Revision 2.
The staff has determined that the material with the:
highest ART at 13 EFPY for Vogtle 1 was the intermediate shell plate B8805-2 with 0.08% copper (Cu), 0.59% nickel (Ni), and an initial RT of 20 F.
The d
materialwiththehighestARTat16EFPYforVogtle2wasthE}owershellplate B8628-1 with 0.05% Cu, 0.59% Ni, and an initial RT f 50 4,
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For the limiting beltline material in Vogtle 1, intermediate shell plate B8805-2, the NRC staff calculated the ART to ba 101.2*F at 1/4T (T = reactor vessel belt -
line thickness and 85.9 F at 3/4T for The staff used a neutron fluence of7.7E18n/cm)at1/4Tand2.7E18n/cm}3EFPY.
at 3/4T.
For the limiting beltline material in Vogtle 2, the NRC staff calculated the' ART forlowershellplateB8628-1tobe111* Fat 1/gTand93.4* Fat 3/4Tfor16EPFY.
2 The staff used a neutron fluence of 9.4E18 n/cm at 1/4T and 3.4E18 n/cm at 3/4T.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 110'F l
at 1/4T and 87 F at 3/4T-for the limiting material for 13 EFPY in Vogtle 1.
Because the licensee's ART of 110'F is more conservative than the staff's ART of 101.2 F, the staff judges that the difference of about 9'F is acceptable.
Substituting the ART of 110'F into equations in 3RP 5.3.2, the staff verified that the proposed Vogtle 1 P/T limits for heatup, cooldown, and hydrotest meet i
the beltline material requirements in Appendix G of 10 CFR Part 50.
1 l-For Vogtle 2, the licensee calculated an ART of 123*F at 1/4T and 97*F at 3/4T for the limiting material for 16 EFPY.
Because the-licensee's ART of 123 F is more conservative than the staff's ART of 111*F, the staff judges that the l
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.o difference of about 12'F is acceptable.
Substituting the ART of 123*F for 1/4T and 97'F for 3/4T into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and hydrotest 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.
l Based on the flange reference temperature of 20'F for Vogtle 1 and 10'F for Vogtle 2, the NRC 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 unirradiated USE in Vogtle 1 is intermediate shell plate B8805-1 with 90 ft-lb.
Using the method inRG1.99,Rev.2,thepredictedCharpyUjEofshellplateB8805-1after being exposed to a fluence of 3.17E19 n/cm (32 EFPY) is 68.4 ft-lb.
The material with the' lowest unirradiated USE in Vogtle 2 is lower shell plate B8825-1 with 83 ft-lb.
Using the method in RG 1.99, Rev. 2, the predicted ChargyUSEofshellplateB8825-1afterbeingexposedtoafluenceof3.17E19 l
l n/cm (32 EFPY) is 63.1 f1-lb.
Both predicted USEs.are greater than 50 ft-lb and, therefore, are acceptable.
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The NRC staff concludes that the proposed P/T limits for the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 13 EFPY for Vogtle 1 and 16 EFPY for Vogtle 2 because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50.
Since the licensee used RG 1.99, Rev. 2, to calculate ART, the proposed P/T limits also satisfy GL 88-11.
Hence, the proposed P/T limits may be incorporated into the Vogtle TSs.
3.0 ENVIRONMENTAL CONSIDERATION
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These amendments involve changes in requirements with respect to the installa-l tion or use of facility components located within the restricted area as defined I
in 10 CFR Part 20 and changes in surveillance requirements.
The staff has determined that the amendments involve 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 si occupational radiation exposure.gnificant increase in individual or cumulative The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding.
Accordingly, the amendments meet 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
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environmental assessment need be prepared in connection with the issuance of the amendments.
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4.0 CONCLUSION
The Consission's proposeo determination that th'e amendments involve no
.i significant hazards consideration was published in the Federal Register on-t November 1, 1990 (54 FR 46149). The Commission consulted with the_5 tate of Georgia.
No public comments.were received, and the State of Georgia did not' l
have any comments. -
The staff has concluded, based on the considerations discussed above that: (1) there is reasonable assurance that the health and safet endangered by operationLin the proposed manner, and-(2)y of the public will not be such activities.will be conducted in compliance with the Commission's regulations, and the issuance of this amendment will nct be inimical to the common defense and security or to the health and safety of-the public.
Prir.cipal Contributor:
.J. Tsao, EMCB Dated:
October 19, 1990 l
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