ML17249A300
| ML17249A300 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 11/13/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | White L ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| TASK-03-08.C, TASK-3-8.C, TASK-RR NUDOCS 7911280008 | |
| Download: ML17249A300 (7) | |
Text
1 TRI'BurION
., Docket n
,NRC PDR Local PDR ORB 82 Reading NRRReading DEisenhuh RHYo1 lmer DLZiemann OELD
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TYMambach HSmitg DCrutchfield (2)
.TERA ACRS (16)
, Docket No. 50-244 Hr. Leon D. tthite, Jr.
Vice President Electric and Steam Producti'on Rochester Gas and Electric Corporation 89 East Avenue Rochester, Nevi York 14649
Dear Nr. Hhite:
RE:
SEP TOPIC III-8 C - IRRADIATION DAtiAt'iEh USE OF SENSITIZED STAINLESS STEEL AND FATNgE RFSISTANCE Enclosed is a copy of our draft evaluation of Systematic Evaluation Program Topic III-B.C.
You are requested to examine the facts upon which the staff has based its evaluation and respond either by confirming that the facts are correct, or by identifying any errors.
If in error, please supply corrected infornation for the docket.
'.,'e encourage you to supply for the docket any other material related to 'these topics that might affect the staff's evaluation.
4 Your response viithin 30 days of the date you receive this letter is requested.
If no response is received within that time, we will assume that you have no comments or corrections.
Enclosure:
, Topic III-8.C cc w/enclosure:
See next page orricc~
OURNAMCW DATO+
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November 13, 1979 cc w/enclosure:
Lex K. Larson, Esquire
- LeBoeuf, Lamb, Leiby & MacRae 1757 N Street, N.
W.
Washington, D.
C.
20036
'r. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Rochester Committee for Scientific Information Robert E. Lee, Ph.D.
P. 0.
Box 5236 River Campus
- Statior, Rochester, New York 14627 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza
- Albany, New York 12223 Director, Technical Development.Programs State of New York Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D.
C.
20555 Or. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D.
C.
20555 Dr.
Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D.
C.
20555 Rochester Publ ic Library 115 South Avenue Rochester, New York 14604 K M C, Inc.
ATTN:
Richard Schaffstall 1747 Pennsylvania
- Avenue, N.
W.
Suite 1050 Washington, D.
C.
20006
SYSTEMATIC EVALUATION PROGRAM PLANT SYSTEMS/MATERIALS R.
E.
CINNA NUCLEAR POWER PLANT Topic III-8.C - Irradiation Dama'ge, Use of Sensitized Stainless Steel and Fatigue Resistance The safety objective of this review is to determine whether the integrity of the internal structures of operating reactors has been degraded through the use of sensitized stainless steel.
The effect of neutron irradiation and fatigue resistance on material of the internal structures was eliminated from the safety objective of Topic III-8.C in memorandum to D.
G. Eisenhut from D. K. Davis and V. S.
Noonan dated December 8, 1978..
The memorandum concluded that operating experience indicated that no significant degradation of the materials of tie reactor internal structures had occurred as a result of either irradiation damage or fatigue resistance.
Furthermore, the Standard Review Plan does not address neutron irradiation nor fatigue resistance of the materials of the reactor internal structures.
Informa.ion for'his assessment was obtained from the Final Facility Description and Safety Analysis Report, Technical Specifications, Inservice Inspection Program for the 1980-1989 Interval, Safety Evaluation Reports to the ACRS, Licensee Event Reports and PWR Nuclear Power Experience for the Ginna Nuclear Power Plant.
Our assessment is based on information in topical reports on sensittzed stainless steel in PWR nuclear steam supply systems and conversations witt materials engineers at Combustion Engineering, Westinghouse and General Electric Company.
The regulatory position is addressed in Section 4.5.2, "Reactor Internals Materials" of the Standard Review Plan.
The areas currently reviewed in the applicant's SAR are materials specification and the controls imposed on the reactor coolant chemistry, fabrication practices and examination and protection procedures.
The materials specification should comply with Section III of the ASME Boiler and Pressure Vessel Code and the fabrication procedures for the components should satisfy the recommendations of Regulatory Guide 1.31, "Control of Ferrite Content in Stainless Steel Weld Metal" and Regulatory Guide 1.44, "Control of the Use of Sensitized Stainless Steel."
The reactor internals are described and analyzed in Section 3.2.3 of the Final Facility Descrip ion and Safety'Analysis Report.
The SAR states that they are designed to withstand the forces due to weight, preload and dynamic loading, vibration, and earthquake acceleration.
The stress values satisfy those prescribed in Section III, ASME Boiler and Pressure Vessel
- Code, 1965 Edition, including Summer 1965 Addendum.
The internals were analvzed for the 6inna nuclear power plant in.a manner. similar to the analyses performed for the Connecticut
- Yankee, San Ono re, SELNI and Zorita reactors.
The internal s.ructures were carefully inspected after hot-functional tests.
November 13, 1979
The structural
- welds, upper core plate inside :upports, the thermal shield attachments to the. core barrel, including all ockwelds on the devices used to lock the bolts, were checked and no malfunctions or abnormalities were found.
The materials used for constructing the reactor internals were identified
.in the Final Facility Description and Safety Analysis Report as Type 304
'tainless'steel with minor quantities of special purpose alloys, such as Inconel 718 and X-750, 17-4 PH alloy in the H-1100 heat treated condition, and Stellite.
The Safety Evaluation states that "the selection of materials for the reactor vessel internals are compatible with the reactor coolant, and have performed satisfactorily."
We concur from our review of the Final Facility Description and Safety Analysis Report that the structural materials specified for the Ginna nuclear power plant have been proven adequate to current standards by extensive tests and satisfactory per-formance.
However, the report has niither detailed the nondestruction examination methods nor the auxiliary materials specifications and procedures used for fabricating the reactor internals.
In the absence o
test data and statements in the Final Facility Description and Safety Analysis Report to show compliance with the recommendations of Regulatory Guide 1.31, "Control of Ferrite Content ir. Stainless Steel,"
and to assure proper control of welding materials and procedures, we assume for this assessment that the reactor internals contained sensitized stainless steel.
A topical report, WCAP-7477-L, "Sensitized Stainless Steel in Wr"tinghouse PWR Nuclear Steam Supply Systems," written by M. A. Golik, March, 1970, was issued to review the nature of sensitized Types 304 and 316 stainless steel in present and future nuclear steam supply systems.
In reviewing the PWR operating experience with the Shippingport, BR-3, Saxton, Yankee
- Rowe, Selni. Connecticut
- Yankee, San Onofre, and Zorita reactors.
The conclusion was reached that
~o Qeneral problems of interqranular or stress corrosion crackina related to sens-stsze3 sfainless steel has been encountered in PWR operating reactors.
This conclusion was discussed with personnel at'Westinghouse and Combustion Engineering who confirmed the conclusion in the report and updated to current PWR operating experience.
The Licensee Event Reports and the PWR Nuclear Power Experience were reviewed for the Ginna nuclear power plant.
None of the events described were related to the use of sensitized stainless s.eel in the fabrication of the reactor internal structures.
The inservice inspection program for the reactor interval structures for the 1980-1989 inspection interval for the R.
E.
Ginna nuclear power plant will be conducted to the requirements of.Section XI, ASM= Boiler and Pressure Vessel
- Code, 1974 Edition, including Summer 1975 Addendum.
The program is in accordance with paragraph (g), Section 50.55a, 10 CFR Par.
50.
Me conclude from our review of the information submitted by the licensee and the operating.information in the Licensee Event Repor.s together with the PMR Nuclear Power Experience that the integrity of the reactor internal structures of the Ginna nuc ear power plant has not been degraded through the use of sensitized stainless steel.
Furthermore, we conclude that the integrity of the internal structures will be assured by an in-service inspection program in accordance with the requiremen.s of paragraph (g), Section 50.55a, 10 CFR Part 50.