ML19093B049
| ML19093B049 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/15/1978 |
| From: | Stallings C Virginia Electric & Power Co (VEPCO) |
| To: | Case E, Reid R Office of Nuclear Reactor Regulation |
| References | |
| Serial No. 081 | |
| Download: ML19093B049 (3) | |
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VIRGINIA ELECTRIC AND PowER CoMPA.NY RICHMOND, VIRGIN IA 23261 February 15, 1978 Mr. Edson G. Case, Acting Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Connnission Washington,.D. C.
20555 Attn:
Mr. Robert W. Reid, Chief Operating Reactors Branch 4
Dear :
Mr. Case:
Serial No. 081 PO&M/JTB:das Docket Nos. 50-280 License No. DPR..;32 This is to confirm and elaborate on information provided to Mr. Mort Fair-tile on February 8, 1978, regarding the fracture toughness of our Unit No. 1 reactor vessel and our survellance program to monitor irradiation effects upon material properties.
In your letter of May 18, 1977 you requested certain information concerning each of our reactor vessels and associated surveillance specimens.
The basis for your request was. that your review of data received from reactor vessel mate-rial programs which indicated that the materials used in reactor vessel fabri-cation may have a wider variation in sensitivity to radiation damage than ori-ginally anticipated, and that some reactor vessels incorporate more than one heat of materials, including weld metals in their beltline regions; however, all these heats may not be included in the reactor vessel material surveillance pro-gram.
You will utilize the data received from affected utilities to determine if the present specimens reasonably represent the limiting materials in the reactor vessel beltline region.
In our letter of January 23, 1978, Serial No. 212B, we forwarded the requested information.
The Westinghouse Electric Corporation, the nuclear steam supply system sup-plier, was contracted to review archive data to provide the information request-ed in your letter.
As a result of this review, Westinghouse provided preliminary notification to us on February 3, which was confirmed on February 6, 1978, that the material contained in our surveillance capsules was not identical to the material contained in the reactor vessel.
We were further informed by Westing-house that the Regulatory Staff was familiar with the generic situation, and in fact prompted your letter of May 18, 1977.
The basic consequence of this new information is that the data regarding material fracture toughness obtained from the survellance capsules which have been removed*and analyzed is not applicable to the reactor vessel from which it was removed. Therefore, the use of data obtained from surveillance capsule sam-ples is not considered valid in evaluating the reactor vessel, so far as meeting the requirements of Appendix G of 10 CFR50 for continued unit operation are con~
cerned.
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e YIROINIA ELECTRIC AND POWER COMPANY TO e
Mr. Edson G. Case Page No. 2 As you are aware, members of the NRC and Vepco staffs have had numerous communica-tions and discussions on reactor vessel toughness, commencing on August 3, 1975, when we notified you in letter No. 631 that, based on Unit No. 1 surveillance capsulesample results, the upper shelf energy of the weld metal was expected to drop below 50 ft-lb.
In light of the new information which we have received, the basis for those discussions were in error and are no longer applicable.
In this regard, the documents previously forwarded to.you should be treated accord-ingly.
Having been made aware of the error by Westinghouse, we have performed an evaluation to ascertain the properties of the materials in our vessels.
Based on this review, the limiting material in the beltline region of the vessel is the circumferential weld seam which joins the intermediate and lower shell courses of the vessel, in particular that part of the weld seam made with weld control No. SA 1585 (Wire heat No. 72445 and Linde 80 Flux Lot No. 8597).
As shown in Table 2 enclosed with our letter of January 23, 1978, the range of copper in Heat No. 72445 is 0.1-7 to 0.25 percent.
Information received from Westinghouse for weld control material No. SA 1585 indicates that the upper shelf energy is 78 ft-lb.
Based on the."data presented in WCAP -7723, the upper shelf energy for the unirradiated weld metal has previously been reported to be 70 ft-lb.
Regulatory Guide (RG) 1.99, Rev. 1 has been used to predict.when the upper shelf energy is estimated to fall below 50 ft-lb.
By using Figure 2 of RG 1.99 the upper shelf energy is predicted to occur for the limiting weld material (0.25%
copper) when the neutron fluence reaches~ 7 x 1018 n/cm2 *. Based on measured fluences obtained from the surveillance capsule that have been analyzed (these measurements remain valid) a fluence of 0.925 n/cm2 at the~ thickness is equi-valent to one(l) effective full power year (EFPY).
At the end of the current Unit No. 1 fuel cycle, approximately April 1, 1978, the total fluence will be equivalent to 3.32 EFPY,_ thus approximately 4 EFPY of operation, or about 5~
years of calendar operation, remains until 50 ft-lb is reached.
We are advised by Westinghouse that the Point Beach Unit No. 1 surveillance capsules contain the same weld wire (heat No. 72445), but a different lot of Linde 80 flux (Lot No. 8504).
Since this surveillance weld material is the same heat number, the copper content should be about the same; therefore, the Point Beach results are representative of the Surry Unit No. 1 material.
The data from the Point Beach Unit No. 1 surv1~llance weld showed a shelf energy of 53 ft-lb after irradiation to 3.50 x 10 n/cm2 and 52 ft-lb after 7.05 x 1018 n/c~2.
These values are in agreement with Figure 2 for 0.25% copper.
Based on this data, the prediction that the Surry Unit No. 1 weld material will not decrease below 50 ft-lb prior to reaching a fluence of~ 7 x 1018 n/cm2 is rea-sonable.
Based on the information presented above, it is predicted that the weld metal in the Surry Unit No. 1 vessel beltline circumferential. weld seam will not decrease below 50 ft-lb prior to seven(-7) EFPY _of operation.
Since the vessel will have accumulated less than 3.4 EFPY at the end of the current fuel cycle, the requirements of Paragraph V.C of Appendix G, 10 CFR50, need not be satisfied until mid-1982.
This inspection would coincide with the ten(lO) year inservice inspection required by Section XI of the ASME Code.
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YIRGIN~ ELECTRIC AND POWER COMPANY TO Mr. Edson G. Case Page No. 3 As a related matter, we are presently planning to remove another surveil-lance capsule from the Unit No. 1 reactor vessel during the next refueling out-age to commence on or about April 1, 1978 in accordance with the Technical Speci-fications.
Since the material property data that would be obtained from this capsule is no longer valid for our vessel, it does not appear.to be necessary to analyze the material specimens.
In accordance with the Technical Specifica-tions, we plan to remove the next schedule surveillance capsule; however, we will defer analyzing the capsule until a resolution of the generic concern is reached.
We would be pleased to discuss this matter with you at your convenience.
cc:
Mr. James P. O'Reilly Very truly yours,
- 6. ]1)1. ~dt4tt: ;1,fY C. M. Stallings Vice President-Power Supply and Production Operations