ML19325F364
| ML19325F364 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 11/15/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19325F358 | List: |
| References | |
| GL-88-11, NUDOCS 8911200229 | |
| Download: ML19325F364 (4) | |
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20666
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f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
1 RELATED TO FACILITY OPERATING LICENSE NO. NPF-12 SOUTH CAROLINA ELECTRIC & GAS COMPANY l
SOUTH CAROLINA PUBLIC SERVICE AUTHORITY
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VIRGIL C. SUPNER NUCLEAR STATION, UNIT NO. 1 DOCKET NO. 50-395
1.0 INTRODUCTION
r In response to Generic Letter 88-11, "NRC Position on Radiation Embrittlement i
of Reactor Yessel Materials and Its Effect on Plant Operations," the South Carolina Electric & Gas Company (the licensee) requested permission to keep the pressure / temperature (P/T) limits in the Virgil C. Summer Nuclear Station, Unit 1, (Summer) Technical Specifications, Section 3.4.
The request was documented in a letter from the licensee dated January 12, 1989.- This request does r.ot change the effectiveress of the P/T limits of 8 effectise full power years (EFPY).
The original P/T limits were developed based on regulatory l
Guide (RG) 1.99, Revision 1, and were compared to the data from surveillance i
capsules V and V.
The request provides up-to-date P/T limits for the operation of the reactor coolant system curing heatup, cooldown, criticality, and i
hydrotes t.
To evaluate the P/T iimits, the staff uses the following NRC regulations and
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guidance: Appendices G and H of 10 CFR Part 50; the ASTM Standards and the l
ASME Code, which are referenced in Apperdices G and H; 10 CFR 50.36(c)(2)
RG 1.99, Rev. 2; Stanoard 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 Technical Specifications for the operation of the plant.
In particular, 10 CFR 50.36(c)(2) reovires that limiting conditions for operation be included in the Technical Specifications.
The P/T limits are among the limiting conditions of operation in the Technical Specifications for dll commercial nuclear plants in the U.S.
Appendices G and H to 10 CFR Part 50 describe specific requirements for f racture toughness and reactor vessel material survtillance that must be considered in setting P/T limits. An acceptable method for constructing the P/T limits is described in SRP I
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Section 5.3.2.
Appendix G to 10 CFR Part 50 specifies fracture toughness and testing requirements for reactor vessel materials in accordance with the ASME
.l Code and, in particular, that the beltline raaterials in the surveillance capsules be tested in accordance with Appenoix H to 10 CFR Part 50. Appendix H, in turn, refers to the ASTM Standards.
These tests define the extent of vessel embrittlement at the time of capsule witheirawal in teims of the increase in reference temperature. Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the ad,)ustea reference temperature (ART) and Charpy upper shelf energy (USE). Generic Letter 68-11 requested that licensees and permittees use the nethods in RG 1.99, Revision 2, to predict the ef fect of neutron irradiation on reactor vessel materials.
This guide cefines the ART as the sum of unirradiated reference temperature, the increase in reference temperature resulting from neutron irradiation, and a margin to account for uncertainties in the prediction nethod.
Appendix H to 10 CFR Part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules f rom 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, welo, and heat-affected-zone (HAZ) materials of the reactor beltline.
1.0 EVALUATION The staff evaluated the effect cf neutron irradiation embrittlement on each beltline material in the Surmer reactor vessel.
The amount of irradiation embrittlernent was calculated in accordance with RG 1.99, kevision 2.
The staff has determined that the material with the highest ART at 8 EFPY has the lower shell plate (C9923-2) with 0.087 copper (Cu), 0.417. nickel (Ni), and an initial RT of 10'F.
nct The licensee has removed two surveillance capsules from the Summer reactor vessel.
The results from capsules V and V were published in Westinghouse reports WCAP-10814 and WCAP-11726, respectively.
Both surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld netal, end HAZ metal.
For the limiting beltline material, plate C9923-2, the staff calculated the ART to be 94'F at 1/4T (T =preactor vessel beltline thickness) for 8 EFPY and a l
fluenceof9.42Ej8n/cm.
The ART at 3/4T was calculated to be 81*F for 8 EFPY and 3.71E18 n/cm.
The ART was determined by Section 1 of RG 1.99, Revision 2, because the limiting material was not in the surveillance capsules.
The licensee used the method in RG 1.99, Revision 2, to calculate an ART of 94.5'F at 1/4T and 81.3'F at 2/4T at 8 EFPY for the same limiting plate material.
Substituting the ART of 94*F into equations in SRP 5.3.2, the staff verified that the proposeo P/T limits for heatup, cooldown, and hydrotest neet the beltline material requirements in Appendix G to 10 CFR Part 50.
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In addition to beltline materials, Appendix G to 10 CFR Part 50 also imposes P/T l
limits b6 sed on the reference temperature for the reactor vessel closure flange materials.
Section IV.2 of Appendix G states that when the pressure exceeds 20%
r of the preservice system hydrostatic test pressure, the temperature of the closure l
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 opera-tien and by 90'F for hydrostatic pressure tests and leak tests.
Based on the flange reference temperature of 10*F, the staff has determined that the proposed i
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 lif e be above 50 ft-lb.
Based on data from the Sunser Final Safety Analysis Report the material with the lowest initial USE is intermediate shell plate (A9154-1),
with 80.5 ft-lb.
Usingthen.ethodinRG1.99,Revisign2 the predicted Charpy USE oftheplatematerialattheendoflife(6.6E19n/cm)wIllbe57ft-lb.
This is ebove 50 ft-lb and, therefore, is acceptable.
3.0 _ CONCLUSION The staff concludes that the current P/T limits for the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 8 EFPY because the limits conform to the requirements of Appendices G and H to 10 CFR Part 50.
The licensee's submittal also satisfies Generic Letter 88-11 because the licensee used the method in RG 1.99, Revision 2 to calculate the i
ART. Since this SE only covers operation through 8 EFPY, the licensee will reed to submit a revised curve to adress operation past the 8 EFPY prior to reaching that level. As the licensee will be pulling a capsule at their next refueling outage (March 1990), the data from that capsule should be available to generate new P/T limits.
4.0 REFERENCES
1.
Regulatory Guide 1.99 Radittion Embrittlement of Reactor Vessel Materials, Revision 2, May 1988 2.
NUREG-0600, Standard Review Plan, Section 5.3.2 Pressure-Temperature Limits 3.
January 12, 1989, Letter from 0. S. Biadham (SCE&G) to USNRC Document Control Desk;
Subject:
V. C. Summer Nuclear Station Response to Generic Letter 88-11 4.
Final Safety Analysis Report for V. C. Summer Nuclear Station 5.
R. S. Boggs, et al., " Analysis of Capsule U from the South Carolina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel Raotation Surveillance Program, WCAP-10814," Westinghouse Electric Corporation, June 1985 6.
D. G. Colburn, et al., " Analysis of Capsule V from the South Carolina Electric and Gas Company Virgil C. Summer Unit 1 Reactor Vessel Radiation Surveillance Program, WCAP-11726 " Westinghouse Electric Corporation, January 1988 Dated: November 15, 1989
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