ML19332C708
| ML19332C708 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 11/17/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19332C706 | List: |
| References | |
| GL-88-11, NUDOCS 8911280447 | |
| Download: ML19332C708 (4) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCtEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.114 TO TACILITY OPERATING LICENSE NO. DPR-36 MAINE YANKEE ATOMIC POWER COMPANY
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MAINE YANKEE ATOMIC POWER STATION DOCKET NO, 50-309 INTRODUCTION In response to Generic letter 88-11. "NPC Position on Radiation Embrittlement of Reactor Vessel Materials and its Effect on Plant Operations," the Maine Yankee Atomic Power Company (the licensee) requested permission to revise the pressure / temperature (P/T) limits in the Maine Yankee Atomic Power Plant Technical Specifications Section 3.4 The request was documented in a letter from the licensee dated December 2, 1988. The proposed P/T limits were developed based on the data from actuel surveillance capsules, The proposed revision provides up-to-date P/T limits for the operation of the reacter coolant system during heatup, cooldown, criticality, and hydrotest.
i To evaluate the P/T limits, the staff uses 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: 10CFR50.36(c)(2);
Regulatory Guide (RG) 1.99, Revision 2; Standard Review Plan (SRP) Section 5.3.2; and Generic Letter 88-11.
4 Eech licensee authorized to operate a nuclear power reactor is required by f.
10 CFR 50.36 to provide Technical Specifications for the operation of the plant.
Inparticuler,10CFR50.36(c)(2)reouiresthatlimitingconditionsofoperation be included in the Technical Specifications. The P/T limits are among the limiting conditions of operation in the Technical Specifications for all commercial a
nuclear plants in the U.S.
Appendices G and H of 10 CFR Part 50 describe J
specific requirements for fracture toughness and reactor vessel material f
surveillance that must be considered in setting P/T limits. An acceptable method for constructing the P/T limits is described in SRP Section 5.3.2.
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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, i
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.
Appendix G also requires the licensee to predict the effects of neutron irradiation on vessel embrittlement by calculating the adiusted reference temperature (ART) and Charpy upper shelf energy (USE).
Generic Letter 88-11 requested that licensees and permittees use the methods in Regulatorv Guide 1.99, Revision ?, to predict the effect of neutron irradit. tion on reactor vessel raterials. This guide defines the ART as the sum of unirradiated reference temperature, the increate in reference temperature resulting from neutron irradiation, and a rnargin to account for urcertainties in the prediction method, kNkh P
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l 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 M fore startup and that they contain test specimens made from plate, weld, and heat-affected-zone (PAZ) materials of the reactor beltline.
2.0 EVALUATION The staff has evaluated the effect of neutron irradiation embrittlement on each beltline material in the Maine Yankee reactor vessel. The amount of neutron irradiation embrittlement was calculated in accordance with RG 1.99, t
Rev. 2.
The staff has determined that the material with the highest ART (most embrittled) at end of life (EOL) was intermediate-to-lower-shell girth weld r
9-?03 with 0.31% copper (Cu) and 0.74% nickel (Ni).
The licensee reported that weld 9-203 had an initial RT of fortbcsurve111erceweld,butuse09 gen-30'FintestsrunatMaineYankee eric initial RT of -56'F obtained ndt for submerced arc welds using Linde 1092 weld flux.
The licensee has removed three surveillance capsules from Paine Yankee. The results from capsule W-263 were published in Battelle-Columbus Deport BCL-585-21, those from capsule A-25 were published in Effects Technology Peport 75-317, and those from capsule A-35 in Westinchouse Report liCAP-9875.
All surveillance capsules contained Cherry impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
For the limiting beltline material, girth weld 9-703, the staff calculated the ART to be 196* F at 1/4 T (T= reactor vessel beltline thickness) for EOL.
The neutron fluence used at 1/4T was 8.76E18 n/cmF. The ART was determined by the least squares extrapolation method using the data from the three Maine Yankee surveillance capsules. The least souares method is described fn Section 2.1 of RG 1.99, Rev. 2.
The licensee used the methnd in RG 1.99, Rev. 2, to calculate the same ART of l
196'F for the limiting weld metal (b203) in the beltline of the Maine Yankee reactor vessel.
Substituting the ART value of 196*F into the equations in SRP j'
5.3.?, the staff verified that the proposed P/T limits for heatup, cooldown, i
and hydrotest meet the beltline material requirements in Appendix G of 10 CFR l
Part 50.
l In addition to beltline materials, Appendix G of 10 CFR Part 50 also imposed j
P/T limits based on the reference temperature for the reactor vessel closure i
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 flance regions highly stressed by the bolt preload must exceed I
the reference temperature of the material in those regions by at least 120'F for norval operation and by 90'F for hydrostatic pressure tests and leak tests. Based on the 'lange reference temperature of 25'F, the staff has determined that the proposed P/T limits satisfy Section IV.? of Appendix G.
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-3 Section IV.B of Appendix G requires that the predicted Charpy USE at the end of life be above 50 ft-1b. Based on data from an accelerated surveillance capsule withdrawn at 4.56 EFPY, the measured Charpy USE is 50 ft-lb for the limiting girth weld (9-203). This is a 53.3% reduction from the unirradiated value of 107 ft-lb. The neutron fluence of this capsule was 8.84E19, which is far higher than the EOL fluence of 1.47E19. Therefore, the USE of the Maine Yankee beltline materials satisfy Section IV.B of Appendix G.
3.0 ENVIRONMENTAL CONSIDERATION
l Notice of Consideration by the staff of issuance of the proposed amendment was i
published in the Federal Register on January 23,1989(54FR3167)andno comments or requests for hearing were received. The Comission also consulted with the State of Maine and no comments were received. An Environmental Assessment (EA) and Finding of No Significant Impact was published in the Federal Register on November 13, 1989 (54 FR 47277). Based upon the EA, the staff has determined not to prepare an environmental impact. statement for the i
proposed license amendment, and has concluded that the proposed action will not have a significant adverse effect on the quality of the human environment.
4.0 CONCLUSION
i The staff concludes that the proposed P/T limits for the reactor coolant system for heatup, cooldown, leak test, and criticality are valid for cumulative thermal generation no greater than 5.414E8 MWH(t), which is equivalent to 22.9 EFPY, because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The licensee's submittal also satisfies Generic Letter 88-11, because the licensee used the method in RG 1.99, Rev. 2 to calculate the ART. Hence, the proposed P/T limits may be incorporated into the Maine Yankee Technical Specifications.
5.0 REFERENCES
1.
Regulatory Guide 1.99, Radiation Embrittlement of Reactor Ves,el Materials, Revision 2, May 1988 2.
NUREG-0800, Standard Review Plan, Section 5.3.2 Pressure-Temperature Limits 3.
"Results of an Assessment of Reactor Pressure Yessel Beltline Materials Required by 10 CFR 50.61 (Pressurized Thermal Shock) for the Maine Yankee Atomic Power Plant " Maine Yankee Atomic Power Company, January 23, 1986 (accessionnumber 8601280169) 4.
Maine Yankee Final Stfety Analysis Report, Section 4.3, " Component and System Design and Operation" 5.
J. W. Sheckhard and R. A. Wu11aert, " Evaluation of the First Maine Yankee Accelerated Suneillance Capsule." CR 75-317. Effects Technology, Inc.,
Santa Barbara, CA, August 15, 1975 6.
J. S. Perrin et al, "Meine Yankee Nuclear Plant Reactor Pressure Vessel Surveillance Program: Capsule 263," BCL-585-21, Battelle-Columbus Lahereteries, Colunius, OH, December 2,1980
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S. E. Yanichko et al, " Analysis of the Maine Yankee Reactor Vessel Seccnd
, Accelerated Surveil 16nce Capsule (WCAP-9875)," Westinghouse Electric i
Corporaticn Pittsburgh, PA, March 1981 8.
J. W. Sheckherd and R. A. Wullaert, "Unirradiated Mechanical Properties of Maine Yankee Nuclear Pressure Vessel Materials," Effects Technology, i
Inc., Santa Barbara, CA, February 1, 1975.
9.
J. B. Randarza letter to USNRC, December 2,1988 (accession number 0812150092) i Principal Contributor: John Tsao Dated-November 17, 1989 f
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