ML19294B385
| ML19294B385 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/05/1980 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| TASK-03-08.C, TASK-3-8.C, TASK-RR NUDOCS 8002280273 | |
| Download: ML19294B385 (6) | |
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February 5.1980 Docket No. 50-155 Mr. David P. Hoffman Nuclear Licensing Administrator Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201
Dear Mr. Hoffman:
RE: SEP TOPIC III-8.C - Irradiation Damage, Use of Sensitized Stainless Steel and Fatigue Resistance.
Enclosed is a copy of our evaluation of Systematic Evaluation Program Topic III-8.C Irradiation Damage, Use of Sensitized Stainless Steel and Fatigue Resistance. This assessment cocpares your facility, as de-scribed in Docket No. 50-155 with the criteria currently used by the regulatory staff for licensing new facilities. Please inform us if your as-built facility differs from the licensing basis assumed in our assessment.
We have discussed this assessment wi. your staff and believe the facts concerning your plant are correct. Therefore, our review of this topic is complete and this evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This topic assess-ment may be revised 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.
Sincerely,
/1 D K
n-Dennis L. Zieminn, Chief Operating Reactors Branch #2 Division of Operating Reactors
Enclosure:
Completed SEP Topic III-8.C cc w/ enclosure:
See next page V
B00gg80
Mr. David P. Hoffman February F. !)80 cc Mr. Paul A. Perry, Secretary U. S. Environmental Protection Consumers Prwer Cog,any Agency 212 West Mf:higan Avenue Federal Activities Branch Jackson, Michigan 49201 Ragion V Office ATTH:
EIS COORDINATOR Judd L. Bacon, Esquire 230 South Dearborn St.eet Consumers Power Company Chicago, Illinois 60604 212 West Michigan Avenue Jackson, Michigan 49201 Herbert Grossman, Esq., Chairman Atomic Safety and Licensing Board Hunton & Williams U. S. Nuclear Regulatory Comission George C. Freeman, Jr., Esquire Washington, D. C.
20555 P. O. Box 1535 Richmond, Virginia 23212 Dr. Oscar H. Paris Atomic Safety and Licensing Board Deter W. Steketee, Esquire U. S. Nuclear Regulatory Comission 505 Peoples Building Washington, D. C.
20555 Grand Rapids, Michigan 49503 Mr. Frederick J. Shon Sheldon, Harmon, Roisman and Weiss Atomic Safety and Licensing Board 1725 I Street, N. W.
U. S. Nuclear Regulatory Commission Suite 506 Washington, D. C.
20555 Washington, D. C.
20006 Big Rock Point Nuclear Power Plant Mr. John O'Neill, II ATTN:
Mr. C. J. Hartman Route 2, Box 44 Plant Superintendent Maple City, Michigan 49664 Charlevoix, Michigan 49720 Charlevoix Public Library Chri sta-Mari a 107 Clinton Street Route 2, Box 108C Charlevoix, Michigan 49720 Charlevoix, Michigan 49720 Chai rman.
County Board of Supervisors Charlevoix County Charlevoix, Michigan 49720 Office of the Governor (2)
Room 1 - Capitol Building Lansing, Michigan 48913 Director, Technical Assessment Division Office of Radiation Programs (AW-459)
U. S. Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460
SYSTEMATIC EVALUATION PROGRAM PLANT SYSTEMS / MATERIALS BIGTORT0hT NUCLEAR 7LTNT 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 materials of the internal structures was eliminated from the safety objective of Topic III-8.C in mamorandun to D. G. Eisenhut from D. K. Davis and V. S. Noonan dated Dec r ber 8, 1978. The memrandum 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 or fatigue.
Furthermore, the Standard Review Plan (Section 4.5.2) does not address neutron irradiation nor fatigue resistance of the materials of the reactor internal structures.
As a result of incidents of intergranular stress corrosion cracking in piping in the BWR system, special study groups were formed by NRC and industry to evaluate the cause, extent and safety implications of the use of sensitized stainless steel in the nuclear steam supply systems. The study groups identified the lacidents with the recirculation system bypass lines, the core spray lines, and the control rod drive return lines.
It was concluded that the problem was caused by a combination of high total stresses, sensitization of the austenitic stainless steel in the heat affected zones of welds, and the relatively high ox3 gen content of the coolant.
The NRC study group recomr. ended an augmented inservice inspection program for stainless steel piping, more stringent monitoring of the leak detection system, modification of plant operating practice, and the use of alternate materials ircune to intergranular stress corrosion cracking. The study group concluded in NUREG-0531, " Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants," that intergranular stress corrosion cracking in piping would be detected prior to unstable crack growth because of the adequacy of the inservice inspection program and the leak detection system.
Reactor operating experience h65 validated the leak-before-break concept of piping integrity, and, it was conc?uded that through-wall cracks in the pipina systems would be detected before they presented a hazard to the health and safety of the public.
The regulatory position on the use of sensitized stainless steel in reactor internal materials is addressed in the Standard Review Plan Section 4.5.2,
" Reactor Internal Materials." 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 Eoiler and Pressure Vessel Code and 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 internal structures are described in Sections 4 and 5 of the Final Hazards Summary Report for the Big Rock Point Nuclear Plant. The internal components were designed to provide support for the fuel and reaintain structural clearances during normal and accident conditions In addition, the internal components provide passageways for the coolant to cool the fuel and means for adequately separating the steam from the
'001 ant water.
Components of the reactor coolant pressure boundary of the Eig Rock Point nuclear Plant were designed, fabricated, inspected and tested to the requirements of Section I and Section VIII of the ASME Boiler and Fressure Vessel Code,1959 Edition, including applicable code case rul ings. 1:here the Code was not applicable, the design was evaluated from the principles described in the U.S. Navy Bureau of Ship Publication,
" Tentative Structural Design Basis for Reactor Pressure Vessels and Directly Associated Components," April,1958.
The primary criteria for material selection for the reactor internal components were the mechanical properties, the material stability and corrosion resistance in the reactor environment. The materials used for the construction of the reactor internals were identified in the Final Hazards Sumnary Report as Type 304 stainless steel, Inconel, and minor quantities of special purpose materials, such as Stellite, Colmonoy, Graphitar, and 17-4 PH alloy. The structural materials identified have proven adequate for reactor internal construction as a result of extensive tests, prior usage, and satisfactory performance.
As a result of the discovery of a leak in the feedwater inlet nozzle of the Lacrosse reactor vessel in October,1969, and in reply to questions from the staff, the licensee in letters dated Septenber 11,
- 970, and January 12, 1971, identified all the furnace sensitized stainless steel components and the maximum calculated levels to which the components would be stressed in service. The reactor internal components were furnace sensitized, but the maximum level of stress intensity did not exceed 90% of the material yield strength (code allowable) at operating temperature.
Experience has shown that at least three elements in combination are necessary to cause cracking in sensitized stainless steel components.
These are material susceptibility, an oxygenated water envirennent, and a threshold total stress. The Big Rock Point Muclear Plant reactor internal components contain sensitized stainless steel in contact with an oxygen saturated water coolant environment. However, the calculated stresses do not exceed :he threshold stress values associated with intergranular stress corrosion cracking. The threshold stress values are near or greater than the 0.2% off-set yield stress at temperature.
Further, in the reactor environment, stress relaxation may occur due to irradiation and temperature effects.
. The Licensee Event Reports and the BUR Nuclear Power _ Experience were revie.ied for the B' Rock Point Nuclear Plant with regard to reactor internal materials problems. The events are sumnarized as follows:
Seginning with the 1955 refueling outage, roller failure was observed in the peripheral control rod blades.
The failure was attributed to severe coolant turbulence in these locations.
Stress corrosion cracking was not a factor.
In a letter of f ay 1972, the staff concluded that this failure did not endanger the health and safety of the public.
Stress corrosion cracking caused the failure of Type 304 stainless steel beryllium-antimony neutron source capsules (1973). An internal pressure build-up of helium-tritium occurred from the n,c< and n. 2n reactions.
The problem was corrected by replacing the stainless steel with Zircaloy capsules.
During the reactor clean-up of beryllium oxide following this failure, the reactor internal components were removed and inspected.
The examination showed neither intergranular stress corr-osion cracking nor evidence of material degradation in the components.
The inlet diffusers became loose from the reactor wall in April, 1979. The shoulder bolts holding the diffusers in place failed as a result of mechanical vibration.
Stress corrosion cracking was not a factor. As no flow blockage occurred, the health and safety of the public was not endangered.
We conclude from our review of the Licensee Event Reports and the SWR Nuclear Power Experience that, even though the stainless steel reactor interaal components were sensitized, incidents of intergranular stress corrt aion cracking were rare and only resulted in the failure of the neutron source capsules. The failure events were detected by the inservice inspection program. At the time of occurrence, the safety consequence was evaluated by the staff and judged to be of minor significance, presenting no hazard to the health and safety of the public.
The inservice inspection program for the reactor internal components is being conducted during the current interval to the requirements of Section XI of the AS!4E Eoiler and pressure Vessel Code,1974 Edition, including Summer 1975 Addenda. The program is in compliance with paragraph (g) of Section 50.55a of 10 CFR Part 50.
It will assure that the integrity of the inuluded ccmponents is maintained during reactor operation.
. 'de conclude from our review that the stainless steel materials in the reactor internal components are sensitized and that there is an increased potential for cracking du' to operation in an oxygen saturated water environ-i..e n t.
Ho'..ever the incid.. cs of stress corrosion cracking are expec ted to be rare because the total stress level in the internal components is relatively low.
In the unlikely event that intergranular stress corrosion cracking should occur, operating experience has demonstrated that cracks in the corc.ponents will be detected by inservice inspection procedures prior to component failures. !!e conclude that the use of scnsitized stainless steel in the reactor int rnal components at the Big Rock Foint i:uclear Plant is not a hazard to the health and safety of the public.