ML20033E383
| ML20033E383 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/23/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20033E379 | List: |
| References | |
| GL-88-11, NUDOCS 9003120589 | |
| Download: ML20033E383 (4) | |
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4 UNITED STATES l
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I NUCLEAR REGULATORY COMMISSION n
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR-REACTOR REGULATION L
RELATED T0_ GENERIC LETTER 88-11 f
NO. DPR-46 NEBRASKA PUBLIC POWER DISTRICT l
COOPER NUCLEAR STATION
1.0 INTRODUCTION
In response to Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Effect on Plant Operations," the Nebraska PublicPowerDistrict(thelicensee)presentedtheexistingpressure/ temperature
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(P/T)limitsintheCooperNuclearStationTechnicalSpecifications,Section 3.6.
The response was documented in a letter from the licensee dated August 3, 1988. The existing P/T limits were developed based on Section 1 of Regulatory Guide (RG)1.99, Revision 2.
The limits provide for the operation of the reactor t
coolant system during heatup, cooldown, criticality, and hydrotest.
-j 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 pendicesGandH;10CFR50.36(c)(2);RG ASMECode,whicharereferencedinAp(SRP)Section5.3.2;andGenericLetter88-11.
2.99, Rev. 2; Standard Review Plant t
Each licensee authorized to operate a nuclear power reactor is required by 10 CFR 50.36 to provide Technical Sp(ecifications for the operation of the plant.
In particular, 10 CFR 50.36 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 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 l.
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 l
reference temperature. Appendix G also requires the licensee to predict the
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/,* i effects of neutron irradiation on vessel enibrittlement by calculating the adjusted reference temperature (ART) and Charpy upper shelf energy (USE).
Gentric 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 from neutron irradiation, and a nargin to account for uncertainties in the prediction method.
Appendix 11 of 10 CFR part 50 requires the licensee to establish a surveillance program to periodically withdraw surveillance capsules from the reactor vessel.
Appendix 11 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 n.ade from plate, weld, and heat-affected-zone (HAZ) materials of the reactor beltline.
2.0 EVALUATION The staff evaluated the effect of neutron irradiation enibrittlement on each beltline material in the Cooper reactor vessel. The amount of irradiation embrittlennnt was calculated in accordance with RG 1.99, Rev. 2.
The staff has determined that the material with the highest ART at 12 EFPY was the lower shell course longitudinal welds (12420/1092) with 0.22% copper (Cu),1.071 nickel (N1), and an initial RT
-22'F.
NDT The licensee has removed one surveillance capsule from Coo >er.
The steff has ascertained that all surveillance capsules contained Ciarpy impact specinens i
and tensile specimens nede from base metal, wtld retal, and HAZ metal.
for the limiting beltline material, lower shell course longitudinal weld (12420/1092), the staff calculated the ART to be 109.2'F at 1/4T (T = reactor vessel beltline thickness) and 285.9'F for 3/4T at 12 E{PY, The staff used a 2
neutron fluence of 5.6E17 n/cm at 1/4T and 2.9E17 n/cn at 3/4T. The ART was determined using Section 1 of RG 1.99, Rev. 2 because only one surveillance
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capsule has been withdrawn so far.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 110'F at 12 EFPY at 1/4T for the same limiting weld metal.
Substituting the ART of 110'F into equations in SRP 5.3.2, the staff verified that the existing 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 materiale, 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
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Lr exceeds 20% of the preservice system hydrostatic test pressure, the temperature i
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 i
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 l
water reactor vessels when water level is within the normal range for power operation and the pressure is less than 20 percent of the pre-service system i
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 reference.
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temperature of 20*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 beltline material with the lowest unirradiated transverse USE is lower shell plate G-2803-3/C2274-2 with 72 ft-lb, which was calculated from a longitudinal USE of 111 ft-lb.
Using the method in RG 1.99, Rev. 2, the predicted Charpy USE of the lower shell plate at the end of life will be 54.7 ft-lb. This is greater than 50 ft-1b and, therefore, is acceptable.
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3.0 CONCLUSION
The staff concludes that the existing P/T limits for the reactor coolant system for heatup, cooldown, leak test, and criticality are valid through 12 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 existing P/T limits may be maintained in the Cooper Technical Specifications.
4.0 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.
August 18, 1977, Letter from J. M. Pilant (NPPD) to D. K. Davis (USNRC),
Subject:
Reactor Vessel Material Surveillance Program: Cooper Nuclear Station 4.
August 3,1988.LetterfromG.A.Trevors(NPPD)toUSNRCDocument.
Control Desk,
Subject:
Response to Generic Letter 88-11. Radiation Embrittiment of Reactor Vessel Materials, Cooper Nuclear Station 5.
February 22, 1988, LetterfromG.Trevors(NPPD)toUSNRCDocument Control Desk,
Subject:
Supplemental Submittal; Proposal Change No. 48 to the Cooper Nuclear Station Technical Specifications
Qr
'..e. <,
r 6.
Final Safety Analysis Report, Cooper Nuclear Station L
7.
July 6,1987, Letter from G. A. Trevors (NPPD). to USNRC Document Control Desk,
Subject:
Reactor Yessel Surveillance Program Principal Contributor:
J. Tsao Dated:
February 23, 1990 l
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