ML20040F341
| ML20040F341 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 01/29/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-03-08.C, TASK-3-8.C, TASK-RR LSO5-82-01-071, LSO5-82-1-71, NUDOCS 8202090097 | |
| Download: ML20040F341 (6) | |
Text
i January 29, 1982 Docket No. 50-213 m
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Mr. W. G. Counsil Vice President Nuclear Ingineering and Operations C
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Connecticut Yankee Atomic Power Co.
Post Office Box 270 9,
Hartford, Connecticut 06101 p
Dear Mr. Counsil:
SUBJECT:
HADD8M NECK - SEP TOPIC III-8.C. " IRRADIATION DAMAGE, USE OF SENSITIZED STAINLESS STEEL AND FATIGUE RESISTANCE."
Reference:
Letter W.G. Counsil to D.L. Ziemann, SEP Topic III-8.C.
dated December 18, 1979.
Enclosed is a copy of our final evaluation of Systenatic Evaluation Program Topic III-8.C. " Irradiation Damage, Use of Sensitized Stainless Steel and Fatigue Resistance".
This assessment compares your facility as described in Docket No. 50-213 with the criteria currently used by the Regulatory staff for licensing new facil-ities. The final evaluation differs from the draft its that it incorporates your comment, has been rewritten into a new format and has been reworded for clarity.
This evaluation will be a basic input to the integrated safety assessment for your facility. This topic assessment may be changed in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated assessment is completed.
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Sincerely, os (o ess W
Dennis M. Crutchfield, Chief 6'6}e[6f Operating Reactors Branct No. 5 Division of Licensing 5.D**"
Enclosure:
As stated 8202090097 820129
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Mr. W. G. Counsil HADDAM NECK Docket No. 50-213 cc Day, Serry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Superintendent Haddam Neck Plant RFD #1 Post Office Box 127E East Hampton, Connecticut 06424 Mr. James R. Himmelwright Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Russell Library 119 Broad Street Mi ddletown,. Connecticut 06457 Board of Selectmen Tcwn Hall
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Haddam, Connecticut 06103 Connecticut Energy Agency ATTN: Ass'istant Director Research and Policy Development Department of Planning and Energy Policy 20 G-and Street Hartford, Connecticut 06106 U. S. Environmental Protection Agency Region I Office ATTN: EIS COORDINATOR JFK Federal Building-Boston, Massachusetts 02203 Resident Inspector Haddam Neck Nuclear Power Station c/o U. S. NRC East Haddam Post Office East Haddam, Connecticut 06423
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SYSTEMATIC EVALUATION PROGRAM TOPIC III-8.C HADDAM NECK TOPIC:
III-8.C, Irradiation Damage, Use of Sensitized Stainless Steel and Fatigue Resistance I;.
INTRODUCTION The reactor internals are designed to support and orient the reactor core l
and control assemblies, provide a flow path for reactor coolant and sup-port in-core instrumentation. The internals are to withstand the forces due to weight, pre-load of fuel assemblies, control rod dynamic loading, vibration, and loss of coolant accident blowdown coincident with earth-quake accelerations.
SEP Topic III-8.C is intended to determine if the integrity of the reactor internal structures has been degraded through'the use of sensitized 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 expert-ence indicated that no significant degradation of the materials of the l
reactor internal structures had occurred as a result of either irradiation damage er fatigue resistance.
Furthermore, the Standard Review Plan does not address neutron irradiation nor fatigue resistance of the materials of l
the structures.
1 II.
REVIEW CRITERIA General Design Criterion 4, " Environmental and Missile Design Bases",
Appendix A,10 CFR Part 50, requires that components be designed to accomo-date the effects of and be compatible with the environmental conditions as-sociated with normal operation, maintenance, testing and postulated accident.
conditions. The use of sensitized stainless steel in the presence of cer-tain environmental conditions can lead to stress corrosion cracking and the eventual loss of structural integrity of the affected component.
III.
RELATED SAFETY TOPICS e
SEP Topics III-8. A and III.8.B evaluate related items such as control rod drive mechanism integrity and loose parts monitoring, respectively.
. IV. REVIEW GUIDELINES The review of the use of sensitized stainless steel in reactor internals-was conducted in accordance with the acceptance criteria of Section 4.5.2,
" Reactor Internal and Core Support Materials", of the Standard. Review' Plan and Regulatory Guides 1.31. " Control of Ferrite Content in Stainless Steel Weld Metal", and 1.44, " Control of the Use of Sensitized Stainless Steel".
The materials specifications requirements were those of Sections II and III of the ASME Boiler and Pressure Vessel Code.
V.
EVALUATION Information for this assessment was obtained from the Facility Description and Safety Analysis, Technical Specifications Safety Evaluation Reports to the ACRS, Licensee Event Reports, and PWR Nuclear Power Experience for the Haddam Neck Plant. Our assessment is based on information in a topical report on the behavior of sensitized stainless steel in PWR nuclear steam supply systems and conversations with materials engineers at Combustion Engineering, Westinghouse and General Electric Company.
Components of the reactor coolant pressure boundary were designed, fabri-cated, and inspected to the requirements of Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code,1962 Edition, including Summer 1963 Addenda plus applicable code cases. The stress analyses performed (not required by Section VIII) and stress intensity limits applied were in compliance with the rules of Tentative Structural Design Basis for Reactor Pressure Vessels and Directly Associated Components (Pressurized Water Cooled Systems) PB-151987 as modified to account for the stress criteria of Section III of the ASME Boiler and Pressure Vessel Code.
The materials used for constructing the reactor internals were identified in the Facility Description and Safety Analyses as Type 304 stainless steel with minor quantities of special purpose alloys, such as, Inconel 718 and X, Type 410 stainless steel, and cobalt-based alloys. The type of materials used was specified in the Westinghouse Equipment Specification, which, in some cases, upgraded or modified the ASME Code requirements.
Insufficient information was included in the Facility Description and Safety Analyses report to ascertain compliance with the recommendations of Regula-tory Guide 1.31, " Control of Ferrite Content in Stainless Steel Weld Metal", [ _
and Regulatory Guide 1.44, " Control of the 'Use of Sensitized Stiinless Steel"
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to assure proper contFoi of welding materials and procedures. Therefore.
we assume for this assessment that the reactor internal structures contained sensitized stainless steel.
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. 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", writ-ten by M. A. Golik, March,1970. The report reviewed the nature of sensitized Types 304 and 316 stainless steel.and the signific' ant ~ factors in the application of sensitized stainless steel in present and future nuclear steam supply systems.
In reviewing the PWR operating experience of the Shippingport, BR-3, Saxton, Yankee Rowe, Selni, Haddam Neck, 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 conclusion in the report and updated cur-rent PWR operating experience.
The operational experience of the Haddam Neck Plant was reviewed in the licensees Event Reports and PWR Nuclear Power Experience. None of the events described were traceable to the use of sensitized stainless steel in the fabrication of the reactor internal structures.
The following information was contained ~in the Safety Evaluation'by the Division of Reactor Licensing dated July 1,1971:
"During the first refueling outage, the applicant performed an in-spection of the reactor internals... that identified the following conditions:
1.
Breakage of a structural element in each of two of the control rod clusters...
3.
Breakage of the six flexure pieces between the top'of the thermal shield and the core sunport barrel.
The licensee has thoroughly investigated the cause of each deficiency and has taken appropriate action in each case.
The breakage of the control rod cluster assemblies was traced to a manufacturing defect.
In each case, the break occurred in a brazed joint which connects a vane, from which control rods are suspended, to a central hub called a spider. The brazed joint in one assembly was examined in detail in a hot cell and found to have no b:aze mater-i ial. inside the joint, probably due to improper cleaning of the joint prior to brazing. The second failed joint had a similar appearance.
It is not feasible to inspect these joints on all control rod assem-blies #or brazing deficiencies but all joints were visually inspected to verify the integrity of the assemblies before the reactor was re-started. Only two such joints have failed in all the operating reatt-ors and there is no evidence that this condition is prevalent in these a s s embl i es."
-4 "The breakage of tha six support' flexure pieces was due to a metal- ~~
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lurgical problem combined with high-cycle' fatigue stress resulting from vibration of the thermal shield. Based on an analysis of the i structural supports for the thermal shield, the licensee concluded that these flexure pieces are not necessary and therefore these pieces were not reinstalled. We concluded that removal of the six support flexure pieces did not present significant hazards.consid-erations not described or implicit in the safety analysis report."
In November,1977, a third vane to hub failure was discovered on a con-trol rod assembly. The mode of failure was concluded to be a poor braze bond between the vane and hub as experienced in the earlier failures.
Subsequent to these failures and similar failures experienced at other operating reactors, the manufacturer has revised its manufacturing and testing processes to assure that a quality braze is obtained at each vane to hub joint.
The inservice inspection program for the reactor internal structure for the current inspection interval 'for Connecticut Yankee will be conducted to the requirements of Section XI, ASME Boiler and Pressure Yessel Code, 1974 Edition, including Summer 1975 Addenda. The program is in accord-ance with paragraph (g), Section 50.55a,10 CFR Part 50.
VI.
CONCLUSIONS We conclude from our review of the information submitted by the Ticensee and the operating information in the Licensee Event Reports together with the PWR Nuclear Power Experience that the integrity of the reactor inter-nal structures for the Haddam Neck 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 in-spection program in accordance with the requirements of paragraph (g),
Section 50.55a,10 CFR Part 50.
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