ML20034G944

Safety Evaluation Supporting Amend 123 to License DPR-19

ML20034G944
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Site:	Dresden
Issue date:	03/03/1993
From:	Office of Nuclear Reactor Regulation
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ML20034G943	List: ML20034G944
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NUDOCS 9303120138
Download: ML20034G944 (3)

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1. o UNITED STATES

'N NUCLEAR REGULATORY COMMISSION i

WASHINGTON, D. C. 20555 r

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

RELATED TO AMENDMENT NO.123 TO FACILITY OPERCfNG LICENSE NO. DPR-19 COMMONWEALTH EDISON COMPANY DRESDEN NUCLEAR POWER STATION. UNIT 2 DOCKET NO. 50-237

1.0 INTRODUCTION

On September 14, 1992, the Commonwealth Edison Company (CECO, the licensee) requested permission to revise the pressure / temperature (P/T) limits in Section 3.6 of the Dresden Nuclear Power Station, Unit 2, Technical Specifications (TS).

The P/T limits were requested for 16 effective full power years (EFPY).

As of July 1,1992, Dresden, Unit 2, has operated to about 12 EFPY.

On July 2, 19c2, the licensee informed the NRC that the P/T limits in the Dresden, Unit 2, TS require revision.

This was determined as a result of the licensee's review of the reactor vessel material data in response to Generic Letter (GL) 92-01.

The licensee's evaluation using Appendix E to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) determined that the reactor vessel has adequate safety margin aminst brittle fracture; however, as an interim measura, the licensee has used the more conservative P/T limits of Dresden Unit 3 for Dresden Unit 2 operation.

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 American Society for Testing and Materials (ASTM) Standards, and the ASME Code, which are referenced in Appendices G and H; 10 CFR 50.36(c)(2); Regulatory Guide (RG) 1.99, Revision 2; Standard Review Plan (SRP) Section 5.3.2; and GL 88-11.

Seneric Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations," recommends RG 1.99, Revision 2, be used in calculating P/T limits, unless the use of different methods can be justified.

I Each licensee authorized tc operate a nuclear power reactor is required by 10 CFR 50.36 to provide TS for the operation of the plant.

In particular, 10 CFR 50.36(c)(2) requires that limiting conditions for operation be included in the TS. The P/T limits are among the limiting conditions for operation in the TS 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.

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I 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 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 adjusted reference temperature (ART). Generic Letter 88-11 requested that licensees use the methods in RG 1.99, Revision 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.

2.0 EVALUATION The licensee selected properties among all beltline materials in Dresden and Quad Cities reactors that would give the highest ART. The licensen used (1) the chemical composition of longitudinal seam weld, PQ-1300, in Dresden Unit 3, (2) the initial nil ductility transition reference temperature (RT,3) of the longitudinal weld in Dresden Unit 2 (Reference 1), and (3) the predicted neutron fluence of Dresden Unit 3 at 16 EFPY.

The. result of this conservative approach is that the limiting material, from which the proposed P/T limits were constructed, has 0.3% copper, 0.33% nickel, and an initial RT,3 of 40 *F.

The neutron fluence used was 1.8E17 n/cm2 at 1/4T (T -

reactor vessel beltline thickness). The licensee calculated a limiting ART of 91 "F at 1/4T which the staff confirmed to be correct using the RG 1.99 method.

Besides reviewing the licensee's ART calculations, the staff also calculated the ART for each beltline material in the Dresden Unit 2 reactor vessel using tne material data in the licensee's surveillance reports and FSAR.

The staff determined that the highest ART is 59.5 *F at 1/4T based on a neutron fluence of 1.2E17 n/cm2 at 16 EFPY (Reference 8).

The limiting beltline material was l

the lower intermediate shell, B4065-1, with 0.23% copper, 0.52% nickel, and an initial RT,3 of 20 *F.

The licensee's ART of 91 *F is more conservative than the staff's ART of 59.5 *F and is acceptable.

Substituting the ART of 91 *F into equations in l

SRP 5.3.2, the staff verified that the proposed P/T limits for heatup, cooldown, and hydrotest meet paragraph IV.A.2 of Appendix G to 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.

Paragraph IV.A.2 of Appendix G states that when the pressure 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 1

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. 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 on the flange reference temperature of 20 *F, the staff has determined that the proposed P/T limits satisfy paragraph IV.A.3 of Appendix G.

The staff concludes that the proposed P/T limits for heatup, cooldown, leak test, and criticality _ are valid through 16 EFPY because the limits conform to paragraphs IV.A.2, IV.A.3, & IV.A.4 of Appendix G to 10 CFR Part 50. The proposed P/T limits also satisfy GLL 88-11 because the licensee used the method l

in RG 1.99, Revision 2, to construct the limits.

Hence, the proposed P/T i

limits may be incorporated into the Dresden, Unit 2, TS.

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3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of the_ amendment. The State official had no comments.

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4.0 ENVIRONMENTAL CONSIDERATION

The imendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no e

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 l

significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (57 FR 55578). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR i

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.

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5.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, l

that:.(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) 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 Contributor: John Tsao I

Date:

March 3, 1993 l

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6.0 REFERENCES

1.

Appendix F of Dresden, Unit 2, Final Safety Analysis Report.

2.

NUREG-0800, Standard Review Plan, Section 5.3.2:

Pressure-Temperature Limits.

3.

Letter from P. L. Piet (Ceco) to T. E. Murley (USNRC), subject:

Application to Amendment to Facility Operating License DPR-19, Appendix A, Technical Specifications; Proposed Amendment to Figure 3.6.1,

" Minimum Reactor Vessel Metal Temperature," September 14, 1992.

4.

Letter from R. Stols (Ceco) to T. E. Murley (USNRC),

Subject:

Application for Amendment to facility Operating Licenses DPR-19, DPR-25, DPR-29, and DPR-30, October 10, 1989.

5.

Letter from R. Stols (CECO) to T. E. Murley (USNRC),

Subject:

Application for Amendment to Facility Operating Licenses DPR-19, DPR-25, DPR-29, and DPR-30, October 23, 1989.

6.

Letter from R. Stols (Ceco) to T. E. Murley (USNRC),

Subject:

Response

to Request for Additional Information, March 23, 1990.

7.

G. F. Rieger and G. H. Henderson, "Dresden Nuclear Power Station Unit One and Unit Two, Mechanical Properties of Irradiated Reactor Vessel Material Surveillance Specimens," NEDC-12585, May 1975.

8.

E. B. Norris, "Dresden Nuclear Power Station Unit 2 Reactor Vessel Irradiation Surveillance Program, Analysis of Capsule 8,"

SWRI 06-6901-002, March 1983.

9.

E. O. Fromm, et al., "Dresden Nuclear Plant Reactor Pressure Vessel Surveillance Program: Unit No. 2 Capsule Basket Assembly No. 5,"

BCL-585-10, May 8, 1979.

10.

J. S. Perrin, et al., "Dresden Nuclear Plant Reactor Pressure Vessel Surveillance Program: Unit No. 2 Neutron Dosimeter Monitor, Unit No. 2 Capsule Basket Assembly No. 2, and Unit No. 3 Capsule Basket Assembly No. 12,"

BCL-585-3, September 15, 1977.

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