ML20126F536
| ML20126F536 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 05/28/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20126F531 | List: |
| References | |
| NUDOCS 8506170522 | |
| Download: ML20126F536 (3) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION n
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WASHINGTON, D. C. 20555 e
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j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPP0PTING AMENDMENT N0.121 TO LICENSE NO. DPR-49 IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331
1.0 INTRODUCTION
In a letter dated January 11, 1985, the Iowa Electric Light and Power Company (the licensee /IELP) requested an amendment to the Technical Specifications for the Duane Arnold Energy Center (DAEC). The amendment proposes (1) to increase the effectiveness of the pressure-temperature operating limits for DAEC to 12 effective full power years (EFPY), (2) to adjust the minimum vessel head bolting stud temperature, and (3) to revise the reactor vessel surveillance capsule withdrawal schedule.
Submitted in the January 11, 1985 letter was a General Electric (GE) Report NEDC-30839, titled "Duane Arnold Energy Center Reactor Pressure Vessel Fracture Toughness Analysis to 10 CFR 50 Appendix G, May 1983," which contained the analysis in support of the proposal.
2.0 EVALUATION Pressure-temperature limits must be calculated in accordance with the requirements of Appendix G,10 CFR 50, which became effective on July 26, 1983. Pressure-temperature limits that are calculated in accordance with the requirements of Appendix G,10 CFR 50 are dependent upon the initial reference temperature defined in the ASME Code,Section XI, (RT limitingmaterialsinthebeltline,nozzlediscontinuities,andbIo)sure for the N
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flange regions of the reactor vessel and the increase in RT resulting from neutron irradiation damage to the limiting beltline makNial.
The DAEC reactor vessel was procured to ASME Code requirements, which did not specify fracture toughness to determine the initial RT for each reactor vessel material. Hence, the initial RT formathIalsinthe closure flange, nozzle discontinuities, and belkkIne regions of the DAEC reactor vessel could not be determined in accordance with the test requirements of the ASME Code. Therefore, the initial RT for these materialsmustbeestimatedfrommaterialtestdatafromNberslmilar materials used for fabrication of the reactor vessels in the nuclear industry.
The licensee, in developing pressure-temperature limit curves, has estimated the initial RT for the limiting closure flange (Shell No. 4)
NOT 8506170522 850528 PDR ADOCM 05000331 P
l l material as 14 F, the initial RT for the limiting nozzle discontinuity (Standby Liquid Control Nozzle NN material as 58'F, and the initial RT NDT for the limiting beltline (Plate Course No.1) material as 40*F.
These values were determined using the available drop weight and Charpy V-notch test data and the GE Analytical Procedure Y1006A006, which is explained in GE Report NEDC-30839. This method of analysis was developed by GE from an evaluation of nuclear industry reactor vessel materials. The procedure was previously approved by the staff in its Safety Evaluation for LaSalle Unit Nos. I and 2 (NUREG-0519, March 1, 1981 and Supplement No. 1 June 1981).
The increase in RT resulting from neutron irradiation damage was estimatedbythe1Mnseeusingtheempiricalrelationshipdocumentedin 1
Regulatory Guide 1.99, Rev.1, April 1977, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." This method of predicting neutron irradiation damage is dependent upon the predicted amount of neutron fluence and the amounts of copper and phosphorus in the limiting beltline material. According to GE Report NEDC-30839, the limiting beltline material is Plate 80673-1, which has.012 percent phosphorus and.15 percent copper. Using flux wire measurements, a two-dimensional computer program (DOT) to solve Boltzman transport equation, and a one-dimensional computer code (SN10) to calculate radial. flux 4
distributions GE Report NEDC-30839 indicates that the peak gutron,3 fluence at the 1/4 T location at the end of life of DAEC is 4.4 X 10 n/cm'
.(E greater than 1MeV).
The staff used the method of calculating pressure-temperature limits in US NRC Standard Review Plan 5.3.2, NUREG-0800, Rev. 1 July 1981 to determine ~
the amount of time that the proposed pressure-temperature limits are effective. Our conclusion is that the proposed pressure-temperature limits i
meet the safety margins of Appendix G,10 CFR 50 for-12 EFPY and may be incorporated into the plant's Technical Specifications.
The minimurn vessel head bolting stud temperature must comply with Appendix G, 10 CFR 50. Appendix G, 10 CFR 50 requires that this minimum boltup temperature comply with the requirements in Appendix G of Section III, Division 1 of the ASME Code. Based on the initial RT for the limiting closure flange material, the proposed minimum bolt prNad temperature 4
meets these requirements, and is acceptable.
The DAEC reactor vessel material surveillance program must comply with i
Appendix H. 10 CFR 50. This requires that the capsule withdrawal schedule comply with ASTM E 185-82. Based on the amount'of predicted neutron irradiation damage, the proposed capsule withdrawal schedule complies with these requirements. Hence, it may be incorporated into the plartt's Technical Specifications.
3.0 ENVIRONMENTAL CONSIDERATION
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This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.
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. 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 be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets 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 environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such 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:
B. Elliot and J. Kim Dated:
May 28, 1985 4
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