ML20077N967
| ML20077N967 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 08/12/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20077N959 | List: |
| References | |
| NUDOCS 9108150176 | |
| Download: ML20077N967 (4) | |
Text
..
p3 Rf 0g O _
d'%
/y UNITED STATES '
2d)g/[f'o.,, T~ ',. E -
NUCLEAR REGULATORY COMMISSION f
WASHINoTON D C. 20566 SAFETY EVALUATION BY THE,0FFICE OF NUCLEAR REACTOR REGULATION
[
RELATED-TO AMENDMENT NO.172 TO FACILITY OPERATING LICENSE NO. DPR-49 4
IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IDWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331
1.0 INTRODUCTION
in responseLto Generic Letter 88-11~, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Effect on Plant Operations," the Iowa Electric Light-and Power Company (the licensee) requested permission to revise the pressure / temperature (P/T) limits in the Duane Arnold Energy Center Technical' Specifications, Section 3.6.
The request was-documented in a latter from the licensee dated December 23, 1989. This revision also changes the P/T limits from 12 to 16 effective full-power--years (EFPY).
To evaluate the P/T limits, the staff used 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; 10 CFR 50.36(c)(2);
Regulatory Guide (RC) 1.99, Revision 2; Standard Review-Plan (SRP) Section 5.3.2; and-Generic Letter 88-11.
I 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 o)eration in the Technical Specifications for all commercial nuclear plants in tle 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
-j 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 9108150176 910812 t.
PDR ADOCK 05000331 1
P PDR
. reference temperature.
Appendix G also requires the licensee to predict the 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 1
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 Duane Arnold reactor vessel.
The amount of irradiation embrittlement was calculated in accordance with RG 1.99, Rev. 2.
The staff determined that the material with the highest ART at 16 EFPY was the lower shell plate B0402-1(1-19) with 0.13% copper (Cu), 0.47% nickel (Ni), and an initial RT o
0%
ndt The licensee has removed one surveillance capsule, capsule 1 at 288', from the Duane Arnold reactor vessel at 6 EFPY. The surveillance results were published in General Electric report NEDC-31166. The surveillance capsule contained Charpy impact specimens and tensile specimens made from base metal, weld metal, and HAZ metal.
The next capsule will be withdrawn at 15 EFPY, For the limiting beltline material, plate B0402-1(1-19), the staff calculated the ART to be 120.9 F at 1/4T (T = reactor vessel beitline thickngss) and 110.8*F at 3/4T.2The staff used a neutron fluence of 1.75E18 n/cm.at 1/4T and 1.02E18 n/cm at 3/4T.
The ART was determined by Section 1 of RG 1.99, Rev. 2, because only one surveillance capsule has been removed from the reactor vessel.
The licensee used the method in RG 1.99, Rev. 2, to calculate an ART of 120 F at 1/4T for the same limiting plate metal. The staff judges that a difference of 0.9'F hatween the licensee's ART of 120 F and the staff's ART of 120.9 F is acceptable.
Substituting the ART of 120.9 F into equations in SRP 5.3.2, the staff verified that the proposed 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 materials, 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 l
l a.
, exceeds 20% of the preservice system hydrostatic test pressure, the temperature 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 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 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 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 un the flange reference temperature of 14*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 of the beltline materials at end of life be above 50 ft-lb. The licensee prov'ided unirradiated USE data for all beltline materials except for lower shell plates C6439(1-18) and B0402(1-19). The limiting (lowest) unirradiated USE for materials With known USEs is 87 ft-lb.
Using the method in RG 1.99, Rev.2, the staff predicts that the limiting USE at E0L will satisfy the 50 ft-lb requirement.
The staff estimates that plates C6439(1-18) and B0402(1-19) need an unirradiated USEs of at least 60 f t-lb and 62 ft-lb, recpectively, in order to satisfy the 50 ft-lb requirement. The staff judges that the USE of these two plates will not fall below the 50 ft-lb requirement for some time because of the relatively low neutron fluence in the Duane Arnold reactor vessel.
However, in order to monitor the USE reduction in the plates, the staff will periodically review the surveillance results from other BWR plants that have similar materials composition as the plates.
The staff concludes that the proposed P/T limits for the reactor coolant system _for heatup,*cooldown, leak test, and criticality are valid through 16 EFPY because the limits conform to the requirements of Appendices G and H of 10 CFR Part 50. The proposed P/T limits also satisfy 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 Duane Arnold Technical Specifications.
O STATE CONSULTATION In accordance with the Commission's regulations, the Iowa State official The State official was notified of the proposed issuance of the amendment.
had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
S This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement. The staff has determined that the amendment involves no significant increase in the amounts, and no significant
4.
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 the amendaent involves no significant hazards consideration and there has been no public comment on such finding (56 FR 31435). Accordingly, the 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 the amendment.
5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that: (1) be endangered by operation in the proposed manner, (2) y of the public will n there is reasonable assurance that the health and safet such activities will.be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
Clyde Y. Shiraki John C. Tsao Date: August 12, 1991
-_