ML19221A044
| ML19221A044 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/10/1976 |
| From: | Maccary R Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7905160523 | |
| Download: ML19221A044 (5) | |
Text
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, gg 0-32h Docket No.
MS 24-12 R. C. DeYoung, Assistant Director for Light Water Reactors Division of Project Management METROPOLITAN EDISON CCMPANY, THREE MILE ISLAND, UNIT NWGER 2 (0L) DOCXET NUMBER S0-320 Plant Name: Three Mile Island, Unit Nun:ber 2 Suppliers: Babcock & Wilcox; Burns & Roe Licensing Stage: OL Docket Number: 50-320 Responsible Branch and Project Manager: LWR 2; H. Silver Task: Revision to SER Review Status: Cocplete Because of new infonnation submitted by the applicant' the Materials Integrity Section of the Materials Engineering Branch, Office of Nuclear Reactor Regulation has revised Section 5.2.3, " Materials," and Section 5.3,
" Reactor Vessel Integrity," of the SER. These revisions will satisfactorily resolve the open ite:n regarding fracture toughness testing of the reactor vessel aterials.
R. R. Maccary, Assistant Director for Engineering Division of Systens Safety Office of Nuclear Reactor Regulation
Enclosure:
Revision to SER cc w/ encl:
II. Potapovs, DSS cc w/o encl:
D. Eisenhut, DOR II. F. Conrad, DSS R. S. Boyd, DPM R. E. Heine: nan, DSS
J. P. Knight, DSS
%2 ticket File (50-320)
K. Kniel, DPM NRR Re= ing File 120 301 S. S. Pawlicki, DSS MTE Eajiing File
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METROPOLITAN EDISCN CCMPANY THREE MILE ISLAND NUCLEAR STATION UNIT NUMBER 2 (OL)
DOCKET NUMBER 50-320 REVISION TO SAFETY EVALUATION MATERIALS ENGINEERING BRANCH MATERIALS INTEGRITY SECTION 5.2.3 Materials Revise the second paragraph to read as follows:
Since this vessel was fabricated prior to the issuance of Appendix G,10 CFR Part 50, all of the tests required by this Appendix were not conducted. Hcwever, from the results of the tests conducted and the conservative estimates cf toughness properties, we concluue that the ferritic materials used for the reactor pressure vessel would meet the requirements of Apper. dix G, 10 CFR Part 50. The fracture toughness tests and procedures required by Section III of the ASME Code, and Appendix G,10 CFR Part 50, for the reactor vessel, provide reasonable assurance that adequate safety margins against the possibility of nonductile behavior or rapidly propagating fracture can be established for the pressure-retaining components of the reactor coolant boundary.
-. 5.3 Reactor Vessel Intecrity Revise this Section to read as follows:
We have reviewed all factors contributing to the structural integrity of the reactor vessel and conclude that there are no "special considerations" that make it necessary to consider potential vessel failure for the Three Mile Island tiuclear Station, Unit tiumber 2.
The design, material, fabrication, inspection, and quality assurance requirements conform to the rules of the ASME Boiler and Pressure Vessel Code, Secticn III,1965 Edition, all Addenda through Sur=er 1967, and all applicable Code Cases. Residual elements were not controlled during manufacture. Copper content of the middle circumferential weld is 0.29 percent which is a relatively high value.
Therefore, there is a possibility that irradiation will cause larger than average increases in the ductile-to-brittle transition temperature of the vessel beltline materials.
Legradation of the toughness properties of these materials will be monitered by the material surveillance prog"am which confor ::s to Apoendix H,10 CFR Part 50 and ASTM E 185-73..The program is described in Topical Report BAW-10100, " Reactor Vessel Material Surveillance Program, "which we have reviewed and found acceptable.
If data from the surveillance 120 303
. program indicate that the toughness properties are marginal, we will require that the vessel be annealed to restore the toughness properties to acceptable values.
Operating limitations on temperature and pressure will be established for this plant in accordance with Appendix G,
" Protection Against Nonductile Failure," of the 1972 Summer Addenda of the ASME Boiler and Pressure Vessel Code,Section III, and Appendix G, 10 CFR Part 50.
The integrity of the reactor vessel is assured because the vessel:
1.
Was designed and fabricated to the high standards of quality required by the ASME Boiler and Pressure Vessel Code and pertinent Code Cascs listed above.
2.
Was made from materials of demonstrated high quality.
3.
Was extensively inspected and tested to provide substantial assurance that the vessel will not fail because of material or fabrication deficiencies.
4.
Will be operated under conditions and procedures and with protective devices that provide assurance that the reactor vessel design conditions will not be exceeded during normal reactor operation or during most upsets in operation, and that the vessel will not fail under the conditions of any of the postulated accidents.
370 304
T 5.
Will be subjected t.v monitoring and periodic inspection to demonstrate that the high initial quality of the reactor vessel has not deteriorated significantly under the service conditions.
6.
May be annealed to restore the material toughness properties if this becomes necessary.
120 305