ML19270F577
| ML19270F577 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/29/1979 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Brian Lee COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 7902150462 | |
| Download: ML19270F577 (5) | |
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UNITED ST ATES
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s JAN 2 91979 Docket Nos.
50-373 and 50-374 Mr. Byron Lee, Jr.
Vice President Commonwealth Edison Company P. O. Box 767 Chicago, Illinois 60690
Dear Mr. Lee:
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO CERTAIN SECTIONS OF APPENDICES G AND H In Amendment No. 40 to your Final Safety Analysis Report, you requested exemptions to certain requirements of Appendices G and H of 10 CFR Part 50.
We have reviewed Amendment No. 40 and conclude that additional information is required for Sections IIIA, II!B, and IVB of Appendix G and Sections IIS and IICl of Appendix H.
We are attaching an enclosed guidance to describe the information required for exemptions in order for us to evaluate those areas of non-compliance.
Please inform us after receipt of this letter of the date you can supply the requested information so that we may factor that date into your review schedule.
Please contact us if you desire any discussions or clarification of the information requested.
Sincerely, b.
lan D. Parr, Chief Light Water Reactors Branch No. 3 Division of Project Management
Enclosure:
As stated cc w/ enclosure:
,See next page 790.2150462
Mr. Byron Lee, Jr. cc: Richard E. Powell, Esq.
Isham, Lincoln & Beale One First National Plaza 2400 Chicago, Illinois 60670 Dean Hansell, Esq.
Assistant Attorney General State of I.llinois 188 West Randolph Street Suite 2315 Chicago, Illinois 60601
ENCLOSURE COMMONWEALTH EDISON CCMPANY LA SALLE COUNTY STATION, UNIT NOS. 1 AND 2 (OL)
DOCKET NUMBERS 50-373/374 We have prepared guidance to describe the infonnation required by the staff to determine whetner exemptions may be granted from compliance to certain specific requieements of Appendices G and H of 10 CFR Part 50.
The applicant shculd ior +.ify the areas of non-compliance and state the reason the requirement is impractical for his facility.
In addition, the method used to provide an adequate degree of conservatism in the analyses of the fracture toughness data in lieu of the specific requirement should be stated in the FSAR.
Justification for non-compliance in the following sections is required for the La Salle County Station, Unit Nos.1 and 2.
Appendix L - Section IIIA This section imposes the requirement of NB-2300 of the ASME Code. Charpy V-notch impact data should be provided on unirradiated test specimens orientated in the transverse and longitudinal directions. Other tests and analytical procedures which may be used in lieu of NB-2300 require-ments (or the requirements of earlier code editions) must have adequate technical justification and supporting data.
Section IIIA should be justified by tabulating the transverse and longitu-dinal Charpy V-notch impact data to demonstrate that the ferritic materials in the pressure boundary comply with 10 CFR Part 50 fracture toughness requirements.
Accendix G - Section IIIB2 The materials used to prepare the test specimens shall be representative of the actual materials of the finished components.
In addition, the requirements of Section IIIC for beltline materials shall be met.
Section IIIC requires that the test specimens shall be taken from excess material and welds following completion of the production longitudinal weld joints, and subject to a heat treatment that produces equivalent metallurgical effects.
A complete description of the specimen preparation, including plate material, weld conditions, and wire and flux material, should be made to demonstrate that the specimens are as near as practical and possess equivalent mechanical properties to the actual materials in the reactor vessel.
. Apoendix G - Section IVA4 Materials for bolting and other fasteners with nominal diameter exceeding 1-inch shall meet the minimum requirements of 25 mils lateral expansion and 45 ft-lbs energy level at the lower of either the preload or the lowest service temperature.
The La Salle test data or other suitable supporting test data and a technical evaluation of tne data should be provided to demonstrate with reasonable assurance that adequate safety margins are provided.
Aopendix G - Section IVB Reactor vessel beltline materials shall have a minimum upper shelf energy of 75 ft-lbs as determined by Charpy V-notch impact testing.
The upper shelf Charpy V-notch impact data should be provided to demonstrate that an adequate margin of safety will exist after loss of fracture toughness properties from irradiation.
In the event that Charpy V-notch upper shelf data were not obtained on the reactor vessel materials, consideration should be given to testing the unirradiated specimens from the surveillance program or oroviding other acceptable supporting data to obtain reliable upper shelf data. The recomendations of Regulatory Guide 1.99 should be referenced for estimating the amount of deterioration as a function of fluence.
Appendix H - Section IIB The surveillance program will comply with the requirements of ASTM E 185-73 as modified by Appendix H.
Justification should be stated in the FSAR to demonstrate that the proposed surveillance program will provide an adequate and acceptable method for detecting change in the fracture toughness properties resulting from irradiation comparable to that obtained by using ASTM E 185-73.
Appendix H - Section IICl The surveillance test specimens shall be taken alongside the fracture toughness specimens required by Appendix G.
A complete description of the specimens used in the surveillance program should be made to demonstrate fracture toughness properties equivalent to
. the properties of the reactor vessel. Justification for the use of the test specimens in lieu of the actual material should be based on a technical evaluation.
Refer also to the requirements of Section IIIB2 of Appendix G.