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OEC. 2 2 M Docket No. STN 50-551 MS 24-12 MEMORANDUM FOR:
D. B. Vassallo, Assistant Director l
for Light Water Reactors Division of Project Management FROM:
'J. P. Knight, Assistant Director for Engineering
SUBJECT:
- A900cK-AttrWIECOP,75TANDARFSAFETY~XNMYSMEPfkT D_ -SAR;j!alOERETaNUMBERicTu un 3g t
Plant Name:
B & W Standard Safety Analysis Report (B-SAR-205)
Supplier: Babcock and Wilcox Licensing Stage: CP Docket Number: STN 50-561 Responsible Branch and Project Manager:
LWR 1; T. H. Cox Requested Completion Date:
December 13, 1976 Reviewer:
G. B. Georgiev Description of Response: Safety Evaluation Review Status: Complete Information submitted by the applicant in the PSAR through Amendment No. 6 has been reviewed by the Metallurgy Section, Materials Engineering Branch. Office of Nuclear Reactor Regulation. Our sections of the Safety Evaluation are enclosed.
Areas not resolved by the applicant are identified as follous:
1.
Lack of additional verification of delta ferrite content of austenitic stainless steel weld filler metal using magnetic measur-ing devices in accordance with the MTEB interim position on Regulatory Guide 1.31.
82011907S1 810403 PDR FOIA MADDEN 80-515 PDR
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. l D. B. Vassallo,
2.
Lack of technical justification for not imposing fracture toughness requirement 3 on ASME Class 2 ferritic materials. We will require that the applicant provides a listing of materials of construction, applicable specification, maximum section thickness and lowest service temperature for each Class 2 components in order that we may perform an evaluation of their adequacy.
j'l J. P. Knight, A'ssistant Director for Engineerir.g i
Division of Systems Safety Office of Nuclear Reactor Regulation
-cc w/ encl:
R. E. Heineman, DSS t
D. Eishnhut, 00R l
J. Miller, NRR J. P. Knight, DSS J. F. Stolz, DPM t
R. J. Bosank, DSS S. S. Pawlicki, DSS i
U. Potapovs, DSS H. F. Conrad, DSS C. D. Sellers, DSS G. B. Georgiev, DSS CC w/o encl:
R. S. Boyd, DPM 4
W. G. Mcdonald, MIPC I
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BABC0CK AND WILC0X B-SAR-205 DOCKET NUMBER STN 50-561 SAFETY EVALUATION MATERIALS ENGINEERING BRANCH METALLURGY SECTION I
R EACTO_R_
Reactor Internals Materials The materials for construction of components of the reactor internals have been identified by specification and found to be in conformance with l
the requirements of Section III of the ASME Code.
s The materials for reactor internals exposed to the reactor coolant have been identified and all of the materials are compatible with the expected environment, as proven by extensive testing and satisfactory performance.
General corrosion on all materials is expected to be negligible.
The controls imposed on reactor coolant chem'istry provide reasonable assurance that the reactor internals will be adequately protected during operation from conditions which could lead to stress corrosion of the materials and loss of component structural integrity.
i The controls imposed upon components constructed of austenitic stainless steel, as used in the reactor internals, satisfy the recommendations of Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel,"
and Regulatory Guide 1.66, " Nondestructive Examination of Tubular Products." The applicant has not fully responded to the NRC Interim position on Regulatory Guide 1.31
" Control of Stainless Steel Welding."
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We will require that verification of the delta ferrite content of weld filler metal used for each welding process be made by determinations on undiluted weld deposits using magnetic measuring devices. Material selection, fabrication practices, examination procedures, and protection procedures performed in accordance with these recomendations provide reasonable assurance the austenitic stainless steel used for reactor internals will be in a metallurgical condition which precludes susceptib-l ility to stress corrosion cracking during service. The use of materials proven to be satisfactory by actual service experience and conformance with the recomendations of these Regulatory Guides constitutes an acceptable basis for meeting in part the requirements of NRC General Design Criteria 1 and 14, Appendix A of 10 CFR Part 50.
Control Rod System St.ructural Ma':erials The mechanical properties of structural materials selected for the control rod system components exposed to the reactor coolant satisfy l
Appendix I of Section III of the ASME Code, or Part A of Section II of the Code, and also the NRC Position that the yield strength of cold I
worked austenitic stainless steel should not exceed 90,000 psi.
i The controls imposed upon the austenitic stainless steel of the system satisfy the recommendations of Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel." The applicant has not fully responded to the NRC Interim Positicn on Regulatory Guide 1.31, " Control i
of Stainless Steel Welding." We will require that verification of the f
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delta ferrite content of weld filler metal used for each welding process be made by determinations on undiluted weld deposits using magnetic measuring devices.
Fabrication and heat treatment practices performed in accordance with these recommendations provide added assurance that l
stress corrosion cracking will not occur during the design life of the i
components.
The compability of all materials used in the control rod system in contact with the reactor coolant satisfies the criteria for Articles NB-2160 and NB-3120 of Section III of the Code.
Both martens-j itic and precipitation-hardening stainless steels have been given tempering or aging treatments in accordance with NRC Positions. Cleaning and cleanliness control are in accordance with ANSI Standard N45.2.1-1973,
" Cleaning of Fluid Systems and Associated Components for Nuclear Power Plants," and Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-cooled Nuclear Power Plant."
Conformance with the codes, standards, and Regulatory Guides indicated above, and with the NRC Positions on the allowable maximum yield strength of cold worked austenitic stainless steel and minimum tempering or aging temperatures of martensitic and precipitation-hardened stainless steels, constitutes an acceptable basis for meeting the requirements of NRC General Design Criterion 26, Appendix A of 10 CFR Part 50.
4 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS i
Material Specifications and Compatibility with Reactor Coolant i
The materials used for construction of components of the reactor coolant l
l pressure boundary (RCPB), including the reactor vessel and its appurten-ances, have been identified by specification and found to be in conformance with the requirements of Section III of the ASME Code.
Special requirements of the applicant with regard to control of residual elements in ferritic materials have been identified and are considered acceptable.
9 The RCPB materials of construction that will be exposed to the reactor coolant have been identified and all of the materials are compatible with the expected environment, as proven by extensive testing and satisfactory performance.
General corrosion of all materials except carbon and low alloy steel will be negligible.
For these materials, l
conservative corrosion allowances have been provided for all exposed surfaces of carbon and low alloy steel in accordance with the require-I ments of the ASME Code,Section III, and the external nonmetallic t
insulation to be used on austenitic stainless steel components conforms t
with the requirements of Regulatory Guide 1.36
" Nonmetallic Thermal Insulation for Austenitic Stainless Steels."
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 have been 1
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shown by tests and service experience to be adequate to protect against corrosion and stress corrosion problems.
The controls imposed on reactor coolant chemistry are in conformance with the recommendations of Regulatory Guide 1.44, " Control of Sensitized Stainless Steel," and provide reasonable assurance that the RCPB components will be adequately protected during operation from conditions t
l that could critically lead to stress corrosion of the materials and loss of structural integrity of a component.
The instrumentation and sampling provisions for monitoring reactor coolant water chemistry provide adequate capability to detect significant changes on a timely basis. The use of materials of proven performance and the conformance with the recommendations of the Regulatory Guides constitutes an acceptable basis for satisfying the requirements of NRC General Design Criteria 14 and 31, Appendix A of 10 CFR Part 50.
Fabrication and Processing of Ferritic Materials Materials selection, toughness requirements, and extent of materials testing proposed by the applicant provide assurance that the ferritic materials used for pressure retaining components of the reactor coolant boundary, including the reactor vessel and its appurtenances, will have adequate toughness under test, normal operation, and transient conditions.
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- t The ferritic materials are specified to meet the toughness requirements of the ASME Code,Section III.
In addition, materials for the reactor vessel are specified to meet the additional test requirements and acceptance criteria of Appendix G,10 CFR Part 50, 1
The fracture toughness tests and procedures required by Section III of the ASME Code, as augmented by Appendix G,10 CFR Part 50, for the reactor vessel, provide reasonable assurances that adequate safety margins t
against the possibility of nonductile behavior or rapidly propagating fracture can be established for all pressure retaining cc.mponents of the reactor coolant boundary.
i The results of the fracture toughness tests to be performed in accordance with the ASME Code and NRC Regulations provide adequate safety margins during operating, testing, maintenance, and postulated accident conditions.
1 Compliance with these Code provisions and NRC Regulations constitutes i
an acceptable basis for satisfying the requirements of NRC General Design Criterion 31, Appendix A of 10 CFR Part 50.
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The controls imposed on welding preheat temperatures and weld cladding i
satisfy the intent of the recommendations of Regulatory Guide 1.50,
" Control of Preheat Temperature for Welding of Low-Alloy Steel," and Regulatory Guide 1.43, " Control of Stainless Steel Weld Cladding of 1
Low-Alloy Steeis." These recommendations provide reasonable assurance
that cracking of components made from low alloy steels will not occur during fabrication, and will minimize the possibility of subsequent cracking due to residual stresses being retained in the weldment.
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The welding procedures used for ferritic steels in limited access areas l
comply with the intent of Regulatory Guide 1.71, " Welder Qualification for Areas of Limited Accessibilit'y." The ultrasonic method for examination of ferritic steel tubular products conform to Regulatory Guide 1.66, " Nondestructive Examination of Tubular Products." The fabrication practices and examination procedures performed in accordance with these recommendations provide reasonable assurance that welds in the reactor coolant pressure boundary (RCPB) will be satisfactory in locations of restricted accessibility, and that unacceptable defects in components of the RCPB will be detected regardless of shape, size or orientation.
Conformance with the Regulatory Guides mentioned constitutes an acceptable basis for meeting the requirements of NRC General Design Criteria 1 and 14, Appendix A of 10 CFR Part 50.
Fabrication and Processing of Austenitic Stainless Steels Within the reactor coolant pressure boundary, no components of austenitic stainless steel have a yield strength exceeding 90,000 psi, in accordance with the NRC Position.
The controls imposed upon components constructed of austenitic stainless steel used in the reactor coolant pressure boundary and for the reactor
J vessel and its appurtenances satisfy the recommendations of Regulatory l
Guide 1.44, " Control of the Use of Sensitized Stainless Steel,"
Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water Cooled Nuclear Power Plants," Regulatory Guide 1.71, " Welder Qualification for Areas of l
Limited Accessibility," and Regulatory Guide 1.66, " Nondestructive 4
Examination of Tubular Products." The applicant has not fully responded to the NRC Interim Position on Regulatory Guide 1.31, " Control Stainless Steel Welding." We will require that verification of the delta ferrite content of weld filler metal used for each welding process be made by l
determinations on undiluted weld deposits using magnetic measuring devices.
Materials selection, fabrication practices, examination procedures, and protection precedures performed in accordance with these recomendations provide reasonable assurance that the austenitic stainless steel in the reactor coolant pressure boundary will be free from hot cracking (microfissures) and in a metallurgical condition which precludes suscept-ibility to stress corrosion cracking during service.
Conformance with the Regulatory Guides and NRC Position mentioned constitutes an acceptable basis for meeting the requirements of NRC General. Design Criteria 1 and 14, Appendix A of 10 CFR Part 50.
i Fracture Toughness of Class 2 and Class 3 Components The applicant has stated that he imposes no fracture toughness requirements on ASME Code Class 2 components. We will require that he provide a listing of the materials of construction, applicable specifications, maximum section thickness and lowest service temperature for each Class 2 component in order that we may perform an evaluation of their adequacy.
Steam Generator Materials l
The materials used in Class 1 and Class 2 components of the steam generators were selected and fabricated according to codes, standards, and specifications acceptable to the staff. The onsite cleaning and cleanliness controls during fabrication conform to the recommendations of Regulatory Guide 1.37, " Cleaning of Fluid Systems and Associated Compon-ents during the Contruction Phase of Nuclear Power Plants." The i
controls placed on secondary coolant chemistry are in agreement with i
established staff technical positions. Conformance with applicable codes, standards, staff positions, and Regulatory Guides constitutes an s
acceptable basis for meeting in part the requirements of General Design Criteria 14, 15, and 31.
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I ENGINEERED SAFETY FEATURES Engineered Safety Features Materials 1
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The mechanical properties of materials selected for the engineered l
safety features satisfy Appendix I of Section III of the ASME Code, or Parts A, B and C of Section II of the Code, and the NRC Position that the yield strength of cold worked stainless steels shall be less than 90,000 psi.
The controls on the pH of the reactor containment sprays and the emergency core cooling water following a postnJated loss-of-coolant accident are adequate to ensure freedom from stress corrosion cracking of the austenitic stainless steel components and welds of the contain-ment spray and emergency core cooling systems throughout the duration of the postulated accident to completion of cleanup. The controls on 4
the use and fabrication of the austenitic stainless steel of the systems satisfy the requirements of Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel." The applicant has not fully responded to the NRC Interim Position on Regulatory Guide 1.31, " Control of Stainless Steel Welding." We will require that verification of the delta ferrite content of weld filler metal used for each welding process be made by determinations on undiluted weld deposits using magnetic measuring devices. Fabrication and heat treatment practices performed in accordance with these requirements provide added assurance that-stress corrosion cracking will not occur during the postulated accident s
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time interval.
The control of the pH of the sprays and cooling water, l
in conjunction with controls on selection of containment materials, are in accordance with Regulatory Guide 1.7, " Control of Combustible Gas f
Concentrations in Containment Following a Loss-of-Coolant Accident," and l
provide assurance that the sprays and cooling water will not give rise l
l to excessive hydrogen gas evolution by corrosion of containment metal or l
cause serious deterioration of the containment. The controls placed on i
concentrations of leachable impurities in nonmetallic thermal insulation used on austenitic stainless steel components of the engineered safety features are in accordance with Regulatory Guide 1.36, " Nonmetallic Thermal Insulation for Austenttic Stainless Steel."
Conformance with the Codes and Regulatory Guides mentioned above, and with the NRC Positions on the allowable maximum yield strength of cold worked austenitic stainless steel, and the minimum level of pH of containment sprays and emergency core cooling water constitute an acceptable basis for meeting applicable requirements of NRC General Design Criteria 35, 38, and 41, Appendix A of 10 CFR Part 50, 1
STEAM AND POWER CONVERSION SYSTEM j
Steam and Feedwater' System Materials i
I The mechanical properties of materials selected for Class 2 and Class 3 components of the steam and feedwater systems will satisfy Appendix I of Section III of the ASME Boiler and Pressure Vessel Code, and Parts A, B i
and C of Section II of the Code. The fracture toughness properties of the ferritic materials will satisfy the requirements of Articles NC-2300 and ND-2300 of Section III of the ASME Code and the NRC staff.
e The controls imposed upon austenitic stainless steel comply with the requirements of Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel." The applicant has not fully responded to the NRC Interim Position on Regulatory Guide 1.31
" Control of Stainless Steel Welding." We will require that verification of the delta ferrite content of weld filler metal used for each welding process be made by determin-ations on undiluted weld deposits using magnetic measuring devices.
Fabrication and heat treatment practices that will be performed in accordance with these requirements provide reasonable assurance that stress corrosion cracking will not occur during the design life of the plant. The controls placed upon concentrations of leachable impurities in nonmetallic thermal insulation used on austenitic stainless steel components of the steam and feedwater systems are in accordance with Regulatory Guide 1.36, " Nonmetallic Thermal Insulation for Austenitic Stainless Steel."
The welding procedures used in limited access areas satisfy the intent of the recommendations of Regulatory Guid's 1.71, " Welder Qualification for Areas of Limited Accessibility." The onsite cleaning and cleaniness controls during fabrication satisfy the positions given in Regulatory Guide 1.37, " Quality Assurance Requirements for Cle.'ing of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants," and j
the requirements of ANSI Standard N45.2-1973, " Cleaning of Fluid Systems I
and Associated Components for Nuclear Power Plants." The precautions taken in controlling and monitoring the preheat and interpass taiperatures during welding of carbon and low alloy steel components meet the intent of the recommendations given in Regulatory Guide 1.50, " Control of Preheat Temperature for Welding Low-Alloy Steel."
Conformance with the codes, standards Regulatory Guides, and NRC Positions -
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mentioned constitutes an acceptable basis for assuring the integrity of steam and feedwater systems, and for meeting the requirements of NRC j
General Design Criterion 1, Appendix A of 10 CFR Part 50.
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INTERFACE REQUIREMENTS BY B-SAR-205 ON THE B0P We.have reviewed the materials interface requirements that B-SAR-205 has imposed on the B0P and conclude that they are sufficient and satisfactory.
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0 Fracture Toughness of Class ~2' Components We have reviewed the requirements for fracture toughness testing and properties that the appifcant will meet to provide assurance that the pressure retaining ferritic materials of Code Class 2 components will have adequate toughness. The ferritic materials are specified to meet the toughness requirements of the ASME Code.
The fracture toughness tests and properties required by the ASME Code I
provide reasonable assurance that safety margins against the possibility of nonductile behavior or rapidly propagating fracture can be estabitshed for the pressure-retaining ferritic materials of Code Class 2 components.
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t MATERIALS ENGINEERING BRANCH METALLURGY SECTION REFERENCES General Federal Register 10 CFR Part 50, Appendix A, " General Design Criteria for Nuclear Plants 6" July 7,1971.
j Federal Register 10 CFR Part 50, 5 50.55a, "NRC Codes and Standard Rules - Applicable Codes, Addenda, and Code Cases 'In Effect' for Components that are Part of the Reactor Coolant Pressure Boundary,"
June 12, 1971.
" Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 2 September 1975.
Material Specifications ASME Boiler and Pressure Vessel Code, St-tion III,1974 Edition, including Summer Addenda 1974:
a.
Paragraph NB-2121:
Permitted Material Specifications b.
Paragraph NB-2122:
Special Requirements Conflicting with Permitted Material Specifications I
c.
Specifications for Materials Listed in Tables 1-1.1.1-1.2, and 1-1.3.
ASME Boiler and Pressure Code,Section II,1974 Edition, including Summer Addenda 1974.
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. l Chemistry of Reactor Coolant NRC Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel," May 1973.
Ferritic Steel l
10 CFR Part 50 - Appendix G. " Fracture Toughness Requirements,," June 1,1973.
I ASME Boiler and Pressure Vessel Code,Section III, through 1974 ' Summer Addenda including App'endix G, " Protection Against Non-Ductile Failure."
ASME Specification, SA 370-71b, " Methods and Definitions for Mechanical Testing of Steel Products," ASME Boiler and Pressure Code Section II, Part A - Ferrous,1974 Edition, including Summer,1974 Addenda.
ASTM E 23-73, " Notched Bar Impact Testing of Metallic Materials," Annual Book of ASTM Standards, Part 31, July 1973.
I ASTM E 208-69, " Standard Method foc Conducting Dropweight Test to Deter-mine Nilductility Transiition Temperature of Ferritic Steels," Annual Book of ASTM Standards, Part 31 July 1973.
NRC Regulatory Guide 1.50, " Control of Preheat Temperature for Welding of Low-Alloy Steel." May 1973.
NRC Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water Cooled Nuclear Power Plants," March 16, 1973.
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' J Austenitic Stainless Steel NRC Technical Position - MTEB 5-1, " Interim Position on Regulatory Guide 1.31, ' Control of Stainless Steel Welding.'"
NRC Regulatory Guide 1.36, " Nonmetallic Thermal Insulation for Austenitic Stainless Steel," February 23, 1973.
l NRC Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water Cooled Nuclear Power Plants," March 16, 1973.
NRC Regulatory Guide 1.43, " Control of Stainless Steel Weld Cladding of Low-Alloy Steel," May 1973.
NRC Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel," May 8, 1973.
NRC Regulatory Guide 1.66, " Nondestructive Examination of Tubular Products," October 1973.
NRC Regulatory Guide 1.71, " Welder Qualification for Limited Accessibility Areas " December 1973.
NRC Regulatory Guide 1.7, " Control of Combusti,ble Gas Concentrations in Containment Following a toss-of-Coolant Accident," March 10, 1971.
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I ASTM A 262-70, Practice E " Copper-Copper Sulfate-Sulfuric Acid Test for Detecting Intergranular Attack in Austenitic Stainless Steel "
Annual Book of ASTM Standards, Part 3. April 1973.
ASTM A 393-63, " Recommended Practice for Conducting Acidified Copper Sulfate Test for Intergranular Attack in Auster.itic Stainless Steel,",
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i Annual Book of ASTM Standards, Part 3,. April 1973.
I ANSI N45.2.1-1973, " Cleaning of Fluid Systems and Associated Components l
for Nuclear Power Plants," Draft 2, Revision 0, November 15, 1973, l
American National Standards Institute.
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