ML19329C504
| ML19329C504 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/27/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19329C500 | List: |
| References | |
| NUDOCS 8002140877 | |
| Download: ML19329C504 (8) | |
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1-O TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT 1 (OL)
DOCKET NO. 50-346 SAFETY EVALUATION MATERIALS ENGINEERING BRANCll Materials Application Section KEACTOR Reactor Vessel Internals General v terial Considerations a
We have reviewed the selection of materials for the reactor vessel internals required for reactor shutdown and adequate core cooling. All materials are compatible with the reactor coolant, and have performed satisfactorily in similar applications. Undue usceptibility to intergranular stress-corrosion cracking has been prevented by avoiding the use of sensitized stainless steel in accordance with methods recommended in Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel."
The use of materials proven to be satisfactory by actual service experience and avoidance of sensitization by methods recc= mended in Regulatory Guide 1.44 provides reasonable assurance that the reactor vessel internals will not be susceptible to f ailure by corrosion or stress-corrosion cracking.
The applicant has described the measures that were taken to ensure that deleterious hot cracking of asutenitic steel welds was prevented. All weld filler metal was of selected composition, and welding processes were controlled to limit heat input and co produce welds with at least 57.
delta ferrite, in confornance with recommendations in Regulatory Guide 0002140l
. 1.31, " Control of Stainless Steel Welding." Following these recommenda-tions provides reasonable assurance that no deleterious hot cracking will be present that could contribute to loss of integrity or functional capa-bility.
REACTOR COOLANT SYSTEM AND CMGECTED SYSTEMS Integrity of Reactor Coolant Pressure Boundary General Material Considerations We have reviewed the materials of construction for the reactor coolant i
pressure boundary to ensure that the possibility of serious corrosion or stress-corrosion is minimized. All materials used are compatibl with the expected envircament, as proven by extensive testing and satisfactory service performance. The applicant has shown that the possibility of intergranular stress-corrosion in austenitic stainless steel used for components of the reactor coolant pressure boundary will be minimized because sensitization was avoided, and adequate precautions were taken to prevent contamination during manufacture, shipping, storage, and construction. The measures to avoid sensitization were in general con-formance with the recommendations of Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless," and included controls on compost-tions, heat treatments, welding processes, and cooling rates.
The use of materials with satisfactory service experience and the high degree of conformance with the recommendations of Rc 21atory Guide 1.44 provide reasonable assurance that austenitic stainless steel components will be compatible with the expected service environments, and the probability of loss of structural integrity is minimized.
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_3 Water Chemistry Control Further protection against corrosion problems will be provided by control of the chemical environment. The composition of the reactor coolant will be controlled; and the proposed maximum contaminant levels, as well as the proposed pH, hydrogen overpressure, and boric acid concentrations, tsve been shown by tests and service experience to be adequate to protect against corrosion and stress-corrosion problems.
The possibility that serious corrosion or stress-corrosion problems would occur in the unlikely event that containment spray system operation is required will c minimized because the pH of the recirculating coolant will be maintained at 7.0 by additions of sodium hydroxide.
The controls on chemical composition that will be ir. posed on the reactor coolant and on the recirculating emergency core cooling water provide reasonable assurance that the reactor coolant boundary materials will be adequately protected from conditions that would lead to loss of integrity from stress-corrosion.
Control of Stainless Steel Welding We have reviewed the controls to prevent hot cracking (fissuring)*of austenitic stainless steel welds. These precautions included control of weld raetal composition and welding processes to ensure at least 5%
delta ferrite content in the weld metal. The methods complied with Section III of the ASNE Code, and were in genersi conform'ance with the reccmmendations of Regulatory Guide 1.31, " Control of Stainless Steel Welding." The use of materials, processes, and test methods that were i
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in accordance with these requirements and recommendations provides reasonable assurance that loss of integrity of austenitic stainless steel welds caused by hot cracking during welding will not occur.
CONTAINMENT HEAT REMOVAL AND ECCS SYSTEMS General Material Considerations (Compatibility with coolant)
We have reviewed the materials selection proposed for the containment heat removal and ECCS systems, in conjunction with the expected chemistry of the cooling and containment spray system water.
The applicant has shown that the use of sensitized stainless steel will be avoided, and that the proposed chemistry will not cause stress-corrosion cracking of austenitic stainless steel under conditions that would be present during accident conditions.
We have concluded that the controls on material and cooling water chemistry proposed will provide reasonable assurance that the integrity of components of these systems will not be impaired by corrosion or stress-corrosion.
(Control of SS Welding)
The applicant has shown that welding of austenitic stainless steel for components of these systems will be controlled to prevent deleterious hot cracking. The proposed control of weld metal composition and welding procedures are in general conformance with the recommendations of Regulatory Guide 1.31, " Control of Stainless Steel Welding," and will provide assurance that loss of function will not result from hot cracking of welds."
!!ATERIALS ENCIt:EERING BRANCl!
R EFERE: C r.T.
peneral Fecieral_ itacister 10 CFR Part 50, Appendix A, " General Design Criteria for Maclear Plants," July 7, 1971.
Federal Regiatir 10 CFR Part 50, 5 50.5 a, "AEC Codes and Standard Rules -
Applicable Codes, Addenda, and Code Cases "In Effect" for Ccmponenen that are part of the Reactor Coolant Pressure Ecundary," June 12, 1971.
" Standard format and Content of Safety Analysis Reports for Nuclear Power Plants," Rev. 1, October 1972.
pr., val n,t, rials Concidara tions Ihterial Soccifications
' ASME Boiler and Pressurc Vessel Code,Section III,1971 Edition plus Addenda through Summer 1973.
(a)
Paragraph NS-2121: Permitted I:ateria: Specifications (b) Paragraph N3-2122:
SheecialncquirementsConflictingwithPermitted Material Specificatiens (c)
Specifications for 'htarials Listed in Tab 1cc l-1.1,1-1.2, and L-1.3.
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. (.henistry of Reactor Coolant AEC Reguistory Cuide 1.56,." Maintenance of Unter Purity in Boiling Uater
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Reactorn," June 1973.
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O List of ACC Aporoved Code Cancs, February 22, 1973.
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, Fracture Tcughne.ss 10 CFR 50 - App 6cdin C, " Fracture Toughness Requirements," June 1,1973.
AS:1E Doiler and Prcesure Vessel Code, Seetion III, 1972 Sucerer Addenda, including Appendin.C, " Protection Against Non-Ductile Failure."
AS'!E Specification, SA-370-71b, " Methods and Definitions for !!echanical
. Testing of Steel Products," AS3!E Uoiler and Pressure Vessel Code, ertion TI, c
l' art A - Perrcus,1971 Edition, Sucrer a::d Uinter,1972 l.ddenda.
ASTM Specification E -203-69, " Standard Method for Conducting Dropucight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels,"
Anaual Ecok of ASTd Standards, Part 31, July 1973.
AST!. Specification E 23-72, "N$tched Ear Impact Testing of >!ctallic Materials," Annual Bo.,k of ASTA Standards, Part 31, July 1973.
Ma terial Surveillance Progracts 10 CFR 50 - Appendin H, "ncactor vm:cel Material Surveill taca Progrant
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R aau _r c c.r.t.3, sune 2,
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ASTM Specification E-185-73, " Surveillance Tests on c,truc tural :ta tericls an Nuclear Reactors," Annual Book of,AST' Standards, Part 30, July 1973.
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b Austenitic Stainless Steel AEC Regulatory Guide 1.31, " Control of Stainless Steel Welding," Revision 1, June 1973.
l AEC Regulatory Guide 1.34, " Control of Electro-Slag Wcld Properties,"
December 28, 1972.
AEC Regulatory Guide 1.36, "Nonnetal14c Thermal Insulation for Austenitic Stainless Steel," February 23, 1973.
AEC Regulatory Cuide 1.43, " Control of Stainless Steel Ueld Cladding of Low-Alloy Steel Components," May 1973.
AEC Rc6ulatory Guide 1.44, " Control of the Use of Sensitized Stainicas Steel," May 8, 1973.
'AEC Regulatory Guide 1.50, " Control 'of Preheat Temperature for Welding of Lou-Alloy Steel," May 1973.
ASTM Specificatica, A-262-70, Practice E, " Copper-Copper Sulfate-Sulfuric Acid Test for Detecting Susceptibility to Intergranular Attack in Stainicus Steels," Annual Book of AST 1 Standards,.Part 3, April 1973.
Puen Flyuheela (1) /,EC Replatcry Guide 1.14, " Reactor Coolant Pump F]yuheel Int 0grity,"
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- PCPn Lea! a:,c Detection Systems (1) AEC 5egulatory Guide 1.45, " Reactor Coolant Pressurc Boundary Leahnge Detection Systems," !!ay 1973.
Inservice Insnection Program (1)
AEC Cuideline Docunient, " Inservice Inspection Requirements for Nuclear Power Plants Constructed with Li:aited Accessibility for Inservice Inspections," January 31, 1969.
(2) ASME Boiler and Pressure Vessel Code,Section XI, 1971 Edition, including Winter 1971, Summer 1971, Winter 1972, and Sum =er 1973 Addenda.
P.%s latory Guide 1.51, " Inservice Inspection of ASME, Class 2 and 3 (3) u Nuclear Power Plant Components," May 1973.
Reactor Vessel Intecrity (1) ASME Boiler and Pressure Vessel Code,Section III, 1971 Editica plus Addenda through Winter 1972.
(2) ASME Soiler and Pressure Vessel Cede,Section XI, 1971 Edition plus Addenda through Uinter 1972.
Containment T enkage Testing (1), 10 CFR 50 - Appendiv. J, "neactor Coatainment Leahage Testitq: f o r '.ca t e r-Cooled rower React; ors," February 14, 1973.
(2) American National Standard 15$1 : 45.4-1972, "Lenkage-Ra te Tcotini; of
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Containment Structures for Unc1 car Reactors," March 16, 1972.
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