ML18044A294
| ML18044A294 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 11/16/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-03-08.C, TASK-3-8.C, TASK-RR NUDOCS 7911270492 | |
| Download: ML18044A294 (7) | |
Text
Docket No. 50-255 Local PDR ORB #2 Reading -
N.RR Readfog-DEi senhut RHVolimer OELD Mr. David Bixel Nuclear Licensing Ad~inistrator Consumers Power Company
.or&E** (3)
DLZiemann
-JSWetmore HSmith DCrutchfield * (2)
NSIC NOV 16 Y919 212 West Michigan Avenue.,
Jackson, Michigan 49201
Dear Mr. s*; xe1:
TERA-ACRS (16)
R[:
SEP. TOPIC III-8.C -.IRRADI-ATION DAMAGE, USE OF SENSITIZED STAINL'ESS
- STEEL AMO FATIGUE' RESISTANCE Enclosed is a copy of our draft eval~ation of Systematic E~aluation Program Topi~ JII-8.C. You are requested to examine the facts.upon which th~ staff has based its evaluation and respond etther by corifirming ~hat the f~cts are correct, of by identifying any errors.
If in error, please supply corrected
. information fo.r the docket.
.l*Je encourage you to supply for the docket.any other material related to these topics that might affect "the staff's ev*aluation.
1 Your--response within 30 days of the date you receive this letter is requested.
If,no r_esponse is received within that time,
~-1e will',assume thQt you have no*
comments or corr~ctions.
Enclosure:
Topic II I-8.C
- cc w/enc 1 osur~:
See next page
.' MllC_l'(>RM 3f8 (Sl-76) Nll<X 0240.
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Sincerely, 11 Qriginal Sizned by Tb.0'13-y;; a Vlaro:bach 11 D.en~;f~.l z;e*mann, Chief Operating Reactors Br*anch fi!2 bivi~ion of Operating Reactors*
.y(IA REGUIATDllY oocl([f FlLE COPY;*. </
I 7911270 4'1*~, *c_,C.}
.&. GOVBANMGNT P~INTING 0,,P'ICI!: tlfl
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- 71e I
- i' Mr. Dav id Bi x e 1 cc w/enclosure:
M. I. Miller, Esqµtre Isham, Lincoln & Beale Su.i te 4200 One First National Plaza C~icago, Illinois 60670 Mr. Paul A. Perry, Secretary Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Judd L.
Bacon~ Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Myron M. Cherry, Esquire Suite 4501 One IBM Plaza Chicago, Illinois 60611 Ms. Mary P. Sinclair Great Lakes Energy Alliance 5711 Summerset Drive Midland, Michigan 48640 Charles Bechhoefer, Esq., Chairman Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. c.
20555
- Dr. George C. Anderson Department of Oceanography University of Washington Seattle, Washington 98195 Dr. *M~ Stanley Livingston 1005 Calle Largo Santa Fe, New Mexico 87501 Kalaraazoo Public Library 315 South Rose Street 0
Kalamazoo, Michigan 49006 K M C, Inc.
ATTN:
Richard Schaffstall
- 1747 Pennsylvania Avenue, N. W.
. Suite 1050 Washington, O. C.
20006 Novenber 16, 1979
SYSTEMATIC EVALUATION PROGRAM PLANT SYSTEMS/MATERIALS PALISADES PLANT.
November 16, 1979 TOPIC III-8.C Irradiation Damage, Use of Sensitized Stainless Steel and Fatigue Resistance The safety.objective of this review is to detennine whether the -integrity of the internal structures of operating reactors has been degraded th~ough 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 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 fatigue resistance. of materi a 1 s of the reactor i nterna 1 structures *.
Infonnation for this assessment was obtained from the FSAR, Technical Specifications (Inservice Inspection Requirement), Safety Evaluation Reports to the ACRS, Licensee Event Reports and PWR Nuclear Power Experience for the Pa'l isades Plant.. Our assessment is based on information in topical reports on sensitized sta.inless steel in PWR nuclear steam supply systems and conversations with materials engineers at Com~ustion Engineering, Westinghouse, and General Electric Company.
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... The regulatory position is addressed in Section 4.5.2, "Reactor Internals
!1a~erials 11 of the Standard Review Plan.
The areas currently reviewed in t'.*ie applicant's SAR are materials specification and the controls imposed on tt.: reactor coolant chemistry, fabrication practices and examination.
and protection procedures. The mqterials specifications should comply with ;ection III of the*ASME Boiler and Pressure Vessel Code and the fatJrication procedures 'in the components should satisfy the recorrmendations 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 Palisades FSAR states that the reactor internal structures were designed to support, align and control the core, and transmit operating loads to the flange of the reactor vesse 1..In the Pa 1 i sades Pl ant, they are designed to perform their function and withstand the forces due to deadweight, handling, system pressure, flow-induced pressure drop, flow impingement, temperature differential, shock and vibration. The structures meet the requirements for Class A components stated in Article 4,Section III of the ASME Boiler and Pressure Vessel Code, 1965 Edition, including Winter 1965 Addendum.
The materials used for the construction of the reactor internals were*
identified in Table 4-14 of :the FSAR as Type 304 stainless steel, nickel-chromium-iron (Inconel) alloy aod Stellite.
N~ither Inconel nor Stellite is included in Table N-421 (Design Stress Intensity Values for Carbon and Alloy Steels Excluding Bol~ipg Material) of Article 4 of the referenced section of the ASME Boiler and Pressure Vessel Code.
However, the properties
.. of these materials are gene>'ally well known and were referenced in earlier Codes and Standards, for example,Section VIII, ASME Boiler and Pressure Vessel Code, 1959 Edition. The Sm values cited for Type 304 stainless steel in Table N-421 of Article 4 are equivalent to the accepted values.
We conclude from our review of the FSAR that the basic structural materials specified for the Palisades reactor internals have been proven adequate to current standards by extensive tests an_d satisfactory performance.
However, the Palisades FSAR has neither detailed the nondestructive examination methods nor the auxiliary material specifications and procedures used for fabricating the reactor internals: It states that weld~ng was conducted in accordance with qualifying standards and requirements of Sections III and IX of the ASME Boiler and Pressure Vessel Code,. 1965 Edition. Verification that welding.was performed in accordance with proper procedures and that welder qualification was properly executed were the responsibility of the Engineering Department of the Bechtel Corporation.
The FSAR further states that in order to assure proper.control of welding, the base material, welding electrodes, weld preparation and fitup, welding current, preheat and interpass temperature, travel, cleaning, and appearance of welds were checked by the inspector. *When postweld heat treatment was specified, the inspector checked the thermocouple locations, temperature charts, and the heat treating conditi.ons.
The inspectors were qualified to evaluate all types of welds, including the nondestructive techniques used for verifications-of weld quality *
.. * \\
t'.
>, The FSAR contains neither test data nor other information to substantiate these statements.
In the absence of test data and state~ents in the FSAR to show compliance with the recorrmendations of Regulatory Guies 1.31, "Control of Ferrite Content in Stainless Stee1 Weld Metal", and l.44, "Control of the Use of Sensitized Stainless Steel", we assume for this assessment that the reactors internal structures contained sensitized stain-less stee.1.
A topical report, WCAP-7477*L, "Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems", written by M. A. Golik, March, 1970, was issued to review the nature of sensitized Types 304 arid 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 S.hippingport, ~R-3, Saxton, Yankee.Rowe, Selni, Connecticut Yankee, San Onofre and Zorita reactors, the con~lusion was reached that no general problems of intertjranular or stress corrosion cracking related to sensitized stainless st~el have*been encountered in PWR operating reactors. This conclusion was discussed with personnel at Westinghouse and Combustion Engineering who confinned 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 reactor internals for the Palisades Plant.
None of the events described were related to the use of sensitized stainless steel in the construction of the reactor internal structures.
J
.. *..... The inservice inspection program to assure integrity of the reactor internals will be conducted to the requirements of Section XI, ASME Boiler and Pressure Vessel Code, 1974 Edition, including Sunvner.1975 Addenda.
This is in accordance with Paragraph (g), Section 50.SSa, 10 CFR Part 50.
We conclude from our review of the information submitted by the licensee and the operating experience described in the Licensee Event Reports together with the PWR Nuclear Power Experience that the integrity of the reactor internal structures for the Palisades Plant has not been degraded through the use of sensitized stainless steel. Furthermore, we conclude that the integrity of the internal structure will be assured by an.
inservice inspection program in conformance with the requirements of r
Section XI of. the ASME Boiler and Pressure Vessel Code, 1974 Edition, including Summer 1975 Addenda.
Thi.s program is in compliance to the requirements of Paragraph (g), Section 50.SSa, 10 CFR Part 50.