ML20058C441
| ML20058C441 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 10/18/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058C434 | List: |
| References | |
| GL-88-11, NUDOCS 9011010244 | |
| Download: ML20058C441 (3) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
'j WASHINGTON, D. C. 20555
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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 51 TO FACILITY OPERATING' LICENSE NO..NPF-62 ILLINOIS POWER COMPANY, ET AL.
CLINTON POWER STATION, UNIT NO. 1
' DOCKET NO. 50-461-1
1.0 INTRODUCTION
In response to Generic Letter 88-11. "NRC Position on Radiation Embrittlement I
of Reactor Vessel Materials and Its Effect on Plant Operations," the Illinois 1
(thelicensee)requestedpermissiontorevisethepressure/
Power Company (P/T)-limits in the Clinton Technical Specifications, Section 3.4.
temperature The request was documented in letters from the licensee' dated December 6, 1988 3
l and April 27,1990. This revision changes -the.P/T: limits-from 20 to 12-effective fu11' power years (EFPY). The proposed P/T limits were based on Regulatory Guide 1.99, Revision 2.
The proposed revision:provides up-to-date
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l P/T limits for the operation of the reactor coolant system.during heatup, cooldown, criticality, and hydrotest.
1 To evaluate the P/T limits,. the staff uses the-following NRC regulations and j
guidance: A)pendices G and H of 10 CFR Part 50;'the ASTM Standards and the ASME Code, w11ch 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.
1 Each licensee authorized to operate a nuclear power reactor-is required by j
10 CFR 50.36 to provide Technical Sp(c)ifications for the operation of-the ec plant.
In particular, 10 CFR 50.36 (2) requires:that limiting conditions-j of operation be included in the Technical Specifications. The P/T limits i
are among the limiting conditions of operation in the Technical Specifications for all commercial nuclear plants in the United States., 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.-
I Appendix G bf 10 CFR Part 50 specifies fracture toughness and-testing f"
recuirements for reactor vessel materials in accordance with the ASME Code anc, 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 1
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 9011010244 901o19 fDR ADOCK 05000461 PDC
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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 unirradiated reference temperature, the increase in reference temperature resulting from 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 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 staff evaluated the effect of neutron irradiation embrittlement on each beltline material in the Clinton 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 12 EFPY is weld 76492 with 0.1% copper (Cu),1.08% nickel (N1), and an initial RT of -30*F.
ndt The licensee has not removed any surveillance capsules from the Clinton reactor vessel because the removal date for the first capsule has not been reached. All surveillance capsules contained Charpy impact specimens and tensile specimens made from base metal, weld-metal, and HAZ metal.
For the limiting beltline material, weld 76492 the staff calculated the ART to be 100.3*F at 1/4T (T = reactor vessel beltline thickness) and {9.6*F fo 3/4T at 12 EFPY The staff used a neutron fluence of 1.85E18 n/cm at 1/4T 2
and 9.4E17 n/cm at 3/4T. The ART was determined per Section 1 of RG 1.99, Rev. 2, because no surveillance capsules have been removed'from the Clinton L
reactor vessel.
The licensee calculated an almost identical ART of 100'F for the same weld 76492 using RG 1.99, Rev. 2.
The staff considers the difference-of 0.3'F insignificant (100.3 vs 100'F)., Substituting the ART of 100.3*F into equations in SRP 5.3.2, the staff verified that the proposed P/T limits for L
heatup, cooldown, and hydrotest meet the beltline material requirements in Appendix G of 10 CFR Part 50.
InadditioNtobeltlinematerials,AppendixGof10CFRPart50alsoimposes
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P/T limits based on the reference temperature for the reactor vessel closure flange materials.Section IV.2 of Ap>endix G states'that when the pressure exceeds 20% of the preservice system lydrostatic test pressure, the temperature of the closure flange regions highly stressed by the bolt preload must exceed l
the reference temperature of the material in those regions by at least 120*F y
for normal operation and by 90'F for hydrostatic pressure tests and leak tests.
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Paragraph IV.A.3 of Appendix G states "an exception may be made for boiling-water reactor vessels when water level is within the normal range for power operation and the pressure is less than 20 percent of the preservice~ 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 temperature of O'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 ~1owest unirradiated Charpy impact upper shelf energy is plate A2758-1 with 67 ft-lb'.
Using Figure 2 of RG 1.99, Rev. 2, the staff calculated that the EOL USE would be 55.3 ft-lb. This is greater than 50 ft-lb and, therefore, is acceptable.
The staff concludes that'the proposed P/T. limits for the reactor coolant i
system for heatup, cooldown, lea L 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 catisfies 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 Clinton Technical Specifications.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a requirement with respect to the instal-lation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or a change to surveillance requirement. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of-any effluents that may l
be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission.has previously
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issued a proposed finding that this amendment' involves no significant hazards l
consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forthin10CFR51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement nor environmental; assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
The staff has concluded, based on.the considerations discussed'above, that:
(1) there -is reasonable assurance that the health and safet will not be endangered by operation in-the proposed manner;y of the public and(2)such fr activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: John Tsao Dated: October 18. 1990 m
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