ML13333A436
| ML13333A436 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/15/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | James Drake Southern California Edison Co |
| References | |
| TASK-03-08.C, TASK-3-8.C, TASK-RR NUDOCS 7911270497 | |
| Download: ML13333A436 (5) | |
Text
e NRC PDR Local PDR Docket No. 50-206
'ORB
- 2 Reading NRR Reading DEisenhut RHVollmer DLZiemann NOY OELD OI&E (3)
PWO'Connor ir. Jaes H. Drake HSmith Vice President DCrutchfield (2)
Southern California Edison Company NSIC 2244 Walnut Grove Avenue TERA Post Office Box 800 A
Rosemead, California 91770 ACRS (16)
Dear Mr. Drake:
RE:
SEP TOPIC III-.C -
IRRADIATION DAIAGE, USE OF SENSITIZED STAINLESS STEEL AND FATIGUE'RESISTANCE Enclosed is a copy of our draft evaluation of Systematic Evaluation Program Topic III-8.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 information for the docket.
We encourage you to supply for the docket any other material related to these topics that might affect the staff's evaluation.
Your response within 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.
Sincerely,
~homas3 V0 Gam~Jbach"'
ennis L. Ziemann, Chief Operating Reactors Branch #2 Division of Operating Reactors
Enclosure:
AM Topic III-8.C cc w/enclosure:
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Mr. James November 15, 1979 cc w/enclosure:
Charles R. Kocher, Assistant General Counsel Sou.thern California Edison Company Post Office Box 800 Rosemead, California 91770 David R. Pigott Samuel B. Casey Chickering & Gregory Three,Embarcadero Center Twenty-Third Floor San Francisco, California 94111 Jack E. Thomas Harry B. Stoehr Sanr Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 U. S. Nuclear Regulatory Commission ATTN: Robert J. Pate P 0. Box 4167 San Clemente, California 92672 Mission Viejo Branch Library 24851 Chrisanta Drive Mission Viejo, California 92676 K M C, Inc.
ATTN:
Richard Schaffstall 1747 Pennsylvania Avenue, N. W.
Suite 1050 Washington, D. C. 20006
November 15, 1979 SYSTEMATIC EVALUATION PROGRAM PLANT SYSTEMS/MATERIALS SAN ONOFRE NUCLEAR GENERP'ING STATION UNIT NO. 1 Topic III-8.C -
Irradiation Damage, 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 the 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 fatiaup resistance of the materials of the reactor internal structures.
Information for this assessment was obtained from the Final Safety Analysis Report, Hazards Analysis Reports, Safety Evaluation Reports to the ACRS, Licensee Event Reports,. and PWR Nuclear Power Experience for the San Onofre Nuclear Generating Station Unit No. 1. Our assessment is based on information in topical reports on the behavior of sensitized stainless steel in PWR nuclear steam supply systems, WAPD-SC-541 "PWR Hazards Summary Report for the Shippingport Reactor," WAPD-PWR-971, "Selection and Application of Materials for the PWR Reactor Plant," and conversations with 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 Boiling 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 in Section 2.2.7 of Final Safety Analysis Report. The functions of the core support structure are to support and orientate the fuel eassemblies, maintain orientation and position of the control rods, and to provide passageway for the reactor coolant. The structure consists of an upper and lower support plate, an upper and lower core support barrel; core barrel, radial support and baffle structure. The internals are supported from the reactor flange.
The Hazards Analysis dated November 1, 1963, concluded that "all components of the primary system of the facility are conventional in design, and the materials and design codes proposed to be used are compatible with the operating conditions expected. Accordingly, we believe the primary system will safely perform its intended functions of cooling the core, transferring heat to the secondary system
-2 and containing the primary coolant and associated radioactive materials."
The components of the reactor coolant pressure boundary were designed, fabricated and inspected to the requirements of Section VIII of the ASME Boiler and Pressure Vessel Code. In addition to the static loads within the primary system, thermal and seismic loads were considered in the design of the components.
The materials used for constructing the reactor internals were identified in the FSAR as Type 304 stainless steel with minor quantities of special purpose alloys, such as Inconel X, Type 410 stainless steel, 17-4 PH, and cobalt-base alloys. The type of materials used was specified in Westinghouse Equipment Specification, which, in some cases, upgraded or modified the ASME code requirements.
Insufficient information was included inthe FSAR to ascertain compliance with the recommendations of Regulatory Guide 1.31, "Control of Ferrite Content in Stainless Steel Weld Metal," Regulatory Guide 1.44, "Control of the Use of Sensitized Stainless Steel," and to assure proper control of welding materials and procedures. Therefore, we assume for this assessment that the reactor internal structures contained sensitized stainless steel.
Justification for the use of sensitized stainless steel in PWR quality coolant water was presented in a topical report WCAP-7477-L, "Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," written by M. A. Golik, March, 1970. The report reviewed the nature of sensitized Types 304 and 316 stainless steel and the significant factors in the application of sensitized 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 no general problems of intergranular or stress corrosion related to sensitized stainless steel have been encountered in PWR operating reactors. This conclusion was discussed with personnel at Westinghouse and Combustion Engineering who confirmed the cenclusion in the report and updated to current PWR operating experience.
The operational experience 6f the San Onofre Nuclear Generating Station Unit No. 1 was reviewed in the Licensee Event Reports and the PWR Nuclear Power Experience. There were two events related to the reactor internal structures and both were not traceable to the use of sensitized stainless steel.
These events are summarized as follows:
- 1) Thermal shield vibration was detected during hot functional testing in late 1969.
Following hot functional testing, the core barrel thermal shield was removed for inspection. Cracks were found in the fillet welds which attach the lug and clevis joint keys to the thermal shield. It was concluded that coolant entering the vessel induced ring-mode vibratory response near the upper edge of the thermal shield causing the damage. Six additional supports were installed between the shield and core barrel to absorb the vibration and minimize deflection.
-3 Following the modification and retesting of the system, all welds were inspected and dye checked.
- 2) Remote visual examination of the reactor internals was conducted as part of a routine inspection program on November 1, 1976. It was observed that four of the six flexure supports located at the top of the thermal shield had failed. Cracks were found at both edges of the lower portion of the "S" shaped flexures and progressed the entire horizontal section. The flexures failure was attributed to fatigue due to flow induced vibration. The failure had no adverse effects on the thermal shield and the reactor internal structures.
In letter dated November 16, 1976, to Region V, NRC, Southern California Edison concluded that, after consultation with Westinghouse, satisfactory operation is possible with or without the flexure supports.
The inservice inspection program for the reactor internal structure for the current inspection internal for the San Onofre Nuclear Generating Station Unit No. l.will be conducted to the requirements of Section XI, ASME Boiler and Pressure Code, 1974 Edition, including Summer 1975 Addendum. The program is in accordance with paragraph (g), Section 50.55a, 10 CFR Part 50.
We conclude from our review of the information submitted by the licensee and the operating information in the Licensee Event Reports together with the PWR Nuclear Power Experience that the integrity of the reactor internal structures for the San Onofre 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 inservice inspection program in accordance with the require ments of paragraph (g), Section 50.55a, 10 CFR Part 50.