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UNITED STATES *.
f fg NUCLEAR REGULATORY. COMMISSION g
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W ASHINGTON. D. C. 20555
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N OCT 4 19PA Docket Nos.
50-424/425 MEMORANDUM FOR:
Thomas M. Novak, Assistant Director for Licensing Division of Licensing FROM:
~ William V. Johnston, Assistant Director Materials, Chemical & Environmental-
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Technalogy Division of Engineering
SUBJECT:
DRAFT SER FOR V0GTLE 1 AND 2, SECTIONS 4.5.1, 4.5.2, 5.2.3, 5.3.1 (MATERIALS & FABRICATION), 5.4.2.1, 6.1.1,.
AND 10.3.6 1
Plant Name:
Georgia Power Co., Vogtle Electric Generating Plant, Units 1 and 2 Suppliers: Westinghouse /Bechtel Licensing Stage:
0L Docket Number:
50-424/425 Reviewer:
D. E. Smith Responsible Branch & Project Manager: LB #4; M. Miller Description of Task: Draft Safety Evaluation Report Date Reviewed: 5/2/84 Review Status:
4.5.1 Confirmatory response by applicant required 4.5.2 Complete 5.2.3 Complete 4
5.3.1 (Materials and Fabrication) Complete 5.4.2.1 Complete 6.1.1 Complete 10.3.6 Complete The Materials Application Section, Materials Engineering Branch, Division of Engineering, has completed its review of Vogtle 1 and 2 FSAR for Sections 4.5.1, 4.5.2, 5.2.3, 5.3.1 (Materials &
Fabrication), 5.4.2.1, 6.1.1, and 10.3.6.
This review was conducted in accordance with the applicable sections of the July 1981 edition of the " Standard Review Plan for the Review of Safety Analysis. Reports for Nuclear Power Plants," (SRP, NUREG-0800).
The' acceptance criteria used were in accordance with the SRP, or as described in the text of the SER.
.Contal:t: 'D. E. Smith X24553
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Thomas M. Novak
Our input to th'e Draft Safety Evaluation Report is attached.
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' ' Wili.ciam V. Johnston, Assistant Director Materials, Chemical & Environmental Technology
. Division of Engineering 4
Attachment:
As stated 1
cc:
R. Vollmer D. Eisenhut E. Adensam W. Johnston.
M. Miller E.. Sullivan S. Pawlicki B. D. Liaw C. Cheng W. Hazelton R. Klecker D. E. Smith
'DISTRIBtTTION:
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DMB - Docket Files A EB. Reading Files M TEB Vogtle Filesi t
.SEE PREVIOUS CONCURRENCE SHEET *
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DE:MTEB D'E:MTEB DE:MTEB -
DE:AD:MCET D. Smith W. Hazelton B. D. Liaw W. Johnston 10/ /84 10/
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Thomas M. Novak
'2-Our input to the Draft Safety Evaluation Report is attached.
William V. Johnston Materials, Chemical & Environmental Technology Division of Engineering
Attachment:
As stated R. Vollmer'-
cc:
D. Eisenhut E. Adensam W. Johnston M. Miller E. Sullivan S. Pawlicki B. D. Liaw C. Che.ng W. Hazelto'n R. Klecker D. E. Smith DISTRIBUTION:
DMB - Docket Files MTEB Reading Files MTEB Vogtle Files 3
jC EB bF:lU B DE:MTE DE:AD:MCET ith.
- p. Hazelton B.D.IMw W. Johnston 1/J./84 10/ p /84 10/ 3 /84 10/
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i ATTACNMENT GEO.RGIAPOWERCOMPAN[
V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 AND.2 DOCKET NO.
50-424/425-
,y DRAFT FINAL SAFETY EVALUATION REPORT MATERIALS ENGINEERING BRANCH MATERIALS APPLICATION SECTION 4.5.1 Control Rod Drive Structural Materials The staff concludes that the control rod drive mechanism structural materials are generally acceptable and meet the requimments of General Design Criteria 1,14, and 26 and 10 CFR Part 50, Section 50.55a.
The applicant must confirm that the yield 'stmngth of austenitic stainless steels in these components does not exceed
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90,000 psi.
I This conclusion is based on the applicant having demonstrated that the properties of materials selected for the control red drive mechanism components exposed to the reactor coolani, satisfy Appendix I i
of Section IIT of the ASME Code, and PartsA, B, and C of Section II e
c.f the.; Code..The applicant should confirm conformance with the staff c,
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. position that the yield strength of cold-worked austenitic stain,less steels doe's not exceed 90l000 psi.
Conformance' to the recomendations of Regulatory Guide 1.85 is discussed in 5.2.1.2.
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The controls imposed upon the ferrite content of austenitic stainless steel filler materials satisfy most of the recommendations of Regulatory Guide 1.31,'" Control of Ferrite Content in Stainless' Steel'Wel'd Metal."
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The applicant's ' alternative' approach o'f using chemical analysirs of the weld metal deposit to determine ferrite content is acceptable to the staff.
The controls imposed upon austenitic stainless steels to reduce sensitization satisfy, to the extent practical, the recommendations of Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless
-Steel." The waiving of testing to show non-sensitization of fittings which do not have inaccessible cavities or chambers that would preclude rapid cooling when water quenched or sprayed has been reyiewed and is acceptable to the staff.
The applicant has confirmed that the tempering temperatures and aging temperatures of heat treatable materials in the control rod drive mechanism are specified to eliminate the su meptibility to stress corrosion cracking in reactor coolant.
The fa'b'rication and
' heat treatment practices performed provide' assurance that stress corrosion cracking will not occur during the design life of the components.
The compatibility of all materials used in the control rod system in contact with the reactor coolant satisfies the criteria of Articles NB-2160 and hB-3120 of Section III of the Code.
Cleaning and cleanliness controls.
areinfaccordance,totheextentpractical,withANSIStandardN 45.2.1-1973,
" Cleaning of Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants," and Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning Fluid Systems and Associated Components 4
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'3-of Water-Cooled Nuclear Power Plants." The applicant's use o'f oxygen and nitrogen saturated water of the operating system's quality and during final flushing and use of controlled disposable materials and ' standard cleaning methods to control contamination levels.of harmfuLelements and their compounds is acceptable to the staff.
4.5.2 Reactor Internals Materials The staff concludes that the materia'Is dsed for the construction of the reactor internal and core support structure are acceptable and meet the
. requirements of General Design Criterion 1 and Section 50.55a of 10 CFR Part 50.
The conclusion is based upon the following considerations:
The1 applicant has met the requirements of GDC 1 and Section 50.55a of 10 CFR Part 50 with respect to assuring that the design, fabrication, and testing of the materials used in the reactor internal and core support structure are of high quality standards and adequate f'r' structural integ-o rity.
The controls imposed upon components constructed of austenitic stainless steel satisfy, to the extent practical, the recommendations of Regulatory Guides 1.31, " Control of Ferrite Content in Stainless Steel Weld Metal" and 1.44, " Control of the Use of Sensitized Stainless Steel."
Where the recommendations of these Regulatorys Guide were not follo.wed, the alternative approaches taken by the applicant have been reviewed by the staff and are acceptable (see 4.5.1).
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The materials used for construction of components of the reactor internal and core supportistructure have b'een identified by specifi-cation and found to be in conformance with the requirements of NG-2000 of Secti'on III'and Pairts B, and C of Section'II.ofsthe t
ASME Code.
Conformance'to the recommendations of Regulatory Guide 1.85,
" Code Case Acceptability ASME Section III Materials" is discussed in 5.2.1.2.
As proven by extensive tests and satisfactory performance, the specified materials are compatible with the expected environment and corrosion is expected to be negligible.
The controls imposed on the reactor coolant chemistry provide reasonable assurance that
-the reactor internal and core support structure will be adequately protected during operation-fr.om conditions which could lead to stress corrosion of the materials and loss of component structural integrity.
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The snaterial selection, fabrication practices, examination and testing procedures, and control practices performed in accordance' to these
' recommendations provide reasonable assurance that the materials used, for the reactor internal and core support structure are in a metal-lurgical condition to preclude inservice deterioration.
Conformance with requirements of the ASME Code and the recommendations of the regulatory guides constitute an acceptable basis for meeting in part requirements of General Design Criterion 1 and Section 50.55a of 10 CFR Part 50.
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The staff c6ncludes that the plant design is acceptabl,e. and meets the requirements of General Design Criteria 1,'4,14,30,a'nd3p.cf
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Appendix A of 10 CFR Part 50; the requirements of Appendices B and G
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of 10 CFR'Part 50; and the requirements of 50.55a of 10 CFR Part 50.
This conclusion is based on the staff's review of the FSAR.
The materials used for construction of components of the RCPB have been identified by specification and'found to be in conform-
-ance with the requirements of Section III of the ASME Code.
Compliance'with the above Code provisions for materials specifi-cations satisfies the quality standards requirements of,GDC 1, GDC'30, and 55.55a.
The materials of construction of the RCPB exposed to the reactor coolant have been identified and all of the materials ard compatible 4
with the primary coolant water, which is chemically controlled in accordance with appropriate technical specifications.
This compati-bility has been proven by extensive testing and satisfactory performance.
This includes conforming to the recommendations of' Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel" or the
. alternative approaches taken by the applicant are acceptable to the staff-(gee 4.5.1).
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Generalcorrosionof'allmaterialsincontactwii.hreactorcoolant is negligible, and accordi.ngly, general corrosion' is not of. concern.
Compatibility of the materials with t'he cool. ant and comp 1,'iance" with the Code provisions sat'isfy the rhquirements of GDC 4. relative to compatibility of comp'onents with~ environmental conditions.
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The materials of construction for the RCPB are compatible with the thermal insulation used in these' areas.
The thermal insulation used on the RCPB is either the reflective stainless steel type or is made of nonmetallic compounded materials that are in conformance with the
-recommendations of Regulatory Guide 1.36, " Nonmetallic Thermal Insulationfor Austenitic Stainless Steels." Conformance with the above recommendations satisfies the requirements of GDC 14 and GDC 31 relative to prevention of failure of the RCPB.
The ferritic steel tubular products and the tubular products fabri-cated from austenitic stainless steel have been found to be acceptable i.
'by non-destructive examina ions in accordance with provisions of the ASME Code,Section III.
Compliance with these Code requirements satisfies the quality standards requirements of GDC 1, GDC 30, and 50.55a.
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The fracture toughness tests required by the ASME Code, augmented by Appendix G, 10 CFR Part 50, provide reasonable assurance that l
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'7-adequate safety margins against nonductile behavior or rapid 1'y propagating fracture can be established for all pressure re.taining components df the reactor coolant pressure boundary. ' The' use iof Appendix G of the ASME Code,Section III, and the results of;f.racture toughness tests performed in accordance with the Code and NRC regulations in establishing safe operatin'g procedures, provide adequate safety margins during operating, testing, maintenance, and postulated accid'ent conditions.
Compliance with these Code provisions and NRC regulations satisfies the requir'ements of GDC 31 and 50.55a regarding prevention of fracture of the RCPB.
The applicant has taken alternative approaches to the Tecommendations of Regulatory Guide 1.50, " Control of Preheat Temperature for Welding Low Alloy Steels." The alternative approaches taken by the applicant are that welding procedures are qualified within the preheat temperature I
range rather than at the minimum preheat temperature, and preheat temperatures are maintained for an extended period of time rather than preheat temperatures maintained until the start of post-weld heat i
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treatment.
The staff concludes that these alternative approaches are adequate to prevent hydrogen cracking (the concern of this regulatory guide) and will not cause other hazards.
Accordingly, the staff accepts these alternative approaches. The controls used provide reasonable, assurance that-cracking of components made from low-alloy steels will not occur during fabrication.
If cracking does occur, the required Code inspections should detect-such flaws.
These cont'rols setisfy
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Regulatory Guid'e 1.3'4, " Control of Electroslag Weld Properties," is not applicable because the electroslag welding process was.not used when fabriciting RCPB components.
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- r The controls imposed on' welding ferritic and austenitic steels under conditions of limited accessibility satisfy, to the extent practical, the recommendations of Regulatory Guide 1.71, " Welder Qualification for Areas of Lim'ited Accessibility." The applicant's contractors maintain close supervisory control of -the' welders and reoccurrence of welding situations in production are adequate to assure that the most skilled welders are used in areas-of limited accessibility.
The staff concludes,' that as such welds are inspected, qualification of the welders making acceptable welds occurs automatically under the Code.
These controls satisfy the quality standards requirements of GDC 1, GDC 50, and 50.55a.
The controls imposed on weld cladding of low-alloy steel components by austenitic stainless steel are in accordance with the recommendations of Regulatory Guide T.43, " Control
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ofStainlessSteelWeldCladdingofLow-AlicySteelComponents."
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-The controls to avoid stress corrosion cracking in reactor coolant pressure boundary components constructed of austenitic stainless steels satisfy, to-the extent practical, the recommendations of Regulatory Guide's l.44, " Control of the Use of Sensitized Stainless Steel," and j
1.37, " Quality Assurance Requirements for Cleaning of Fluid Systems
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. acceptable (see 4.5.1).
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The controls followed during material selection, fabrication, examination, protection, sensitization, and contamination, provide reasonable assurance that the RCPB components of austenitic stainless steels are in a ' metallurgical condition that minimizes suscepti-bility to stress corrosion cracking during service. These controls meet the requirements of GDC 4 relative to compatibility of components
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-with environmental conditions and requirements of GDC 14 relative to prevention of leakage and failure of the RCPB.
The' controls imposed during welding of austenitic stainless steels in the RCPB satisfy, to the extent practical, the recommendations of Regulatory Guides 1.31, 1.34 and 1.71 or the alternate approaches taken by the applicant were reviewed by the staff and are' acceptable (see 4.5.1 for Regulatory Guide 1.31, and this section for Regulatory Guides 1.34 and 1.71).
The controls provide reasonable assurance that welded components of austenitic stainless steel did not develop'microfissures during welding
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and have high structural integrity. These controls meet the quality standards requirements of GDC 1, GDC 30, and 50.55a and satisfy the requirements of GDC 14 -relative t'o preiteiitio'n of 'feaka'ge a'nd failure ~
of the RCPB.
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5.3.1 Reactor Vessel Materials (Matefinis and Fabrication) '
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The) staff'c6ncludes that the reactor vessel. material's,are acceptable and t
meettherequirementsOfGehralDesig'nCriteria.1,.4,14,..30,31,and
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32 of Appendix A of 10 C[R Part 50; the material testing and monitoring requirements of Appendices-B, G, and il of 10 CFR Part 50; and the requirements of 50.55a of 10 CFR Part 50.
This conclusion is based on the following:
The materials used for construction of the reactor vessel and its appurtenances ~have been identified by specification and found to be i
in conformance with Section III of the ASME Cbde.
Special require-ments of the applicant with regards to control of residu,a1 elements have been identified and are considered acceptable.
Compliance
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with the above Code provisions for material specifications satisfies the quality standards requirements of GDC 1, GDC 30, and 50.55a.
Ordinary processes seri us'ed for tfie manufacture, fabrication, i
welding, and nondestruci.ive examinations of the reactor vessel and its appurtenances.
Nondestructive examinations in addition to Code fpquirements were also' performed.
Since certification has been 4
made by the applicant that the requirements of Section III of the i
ASME Code have been complied with, the processes and examinations
-r i-used are condsidered acceptable.
Compliance with these Code provisions meets the quality' standards requirements of GDC 1, GDC 30, and 50.55a.
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When welding co'mponents of ferritic steels, Code controls are supplemented by conformance with the recommendations of regulatory guides as'follows:
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The controls imposed on welding preheat temperatures are in I
comformance to the extent practical with the recommendations of Regulatory Guide 1.50, " Control of Preheat Temperature forWeldingofLo'w-AlloySteel." The alternative approaches taken by the applicant were reviewed by the staff and are acceptable (see 5.2.3).
These controls provide reasonable
, assurance that cracking of components made from low-alloy steels did not occur during fabrication and minimize the potential for subsequent cracking.
These controls also satisfy the quality standards requirements of GDC 1, GDC 30, and 50.55a.
b.
Regulatory Guide 1.34, " Control of Electroslag Weld Properties,"
is not applicable because this process is not.used in reactor vessel fabrication.
c.
The controls imposed during weld cladding of ferritic steel components are in conformance with the recommendations of W Regulatory Guide 1.43, " Control of Stainless Steel Weld
' Cladding of Low-Alloy Steel Components." These controls provide -assurance -that-un'derclad 'cr'acking 'did 'not' occur
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during weld cladding of the reactor vessel.
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are suppleme' ted by conformance with ths' recommeridstions of regulatory n
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guides as follows:
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The controls imp'esed ordelta, ferrite in mustenitic stainless a.
steel welds satisfy 'most of the recommendations of Regulatory Guide 1.31, " Control of Ferrite Content in Stainless Steel Weld Metal." The alternate approaches taken by,the applicant have been reviewed by the staff an'd are acceptable (see 4.5.1).
b.
Regualtory Guide 1.34, " Control of Electroslag Weld Properties" is not applicable because this process is not>used in reactor vessel fabrication.
The controls (during, all stages of welding) to avoid contamination s
and sensitization that could cause stress corrosion cracking in austenitic stainless steels conform with the recommendations
- of regulatory guides as follows:
a.
The controls to avoid contamination and excessive sensitiza-tion of austenitic stainless steel satisfy, to the extent
.; practical, the recommendations of Regulatory Guide 1.44,
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l T " Control of the se of Sensitized Stainless Steel." The alternative approaches taken by the applicant have been
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reviewed by the staff and are acceptable (see 4.5.1).'
The controls used provide assurance that welded components weri not' contaminated or excessively. sensitiied prior "to and during the welding process. 'These controls satisfy the quality standards requirement of GDC 1, GDC 30, and 50.55a and the GDC 4 requirement relative to material compatibility.-
b.
The controls regarding onsite cleaning and cleanliness controls of austenitic stainless steel aN in conformance with the recommendations of Regulatory Guide 1.37, " Quality Assurance
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Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants" or the applicant's alternative approaches are acceptable to the staff as discussed in 4.5.1.
These controls provide assurance that austenitic stainless steel components were properly cleaned onsite and satisfy Appendix B of 10 CFR Part 50 regarding controls for onsite cleaning of materials and components.
Integrity of the reactor vessel studs and fasteners is assured by conformance to the extent practical with the recommendations of Regulatory G0ide 1.65, " Materials and Insp'ections for Reactor Vessel
-r Closure S,tuds." The applicants alternative approaches of (a), using a modifieJ SA-540, Grade B-24 for closure stud material which is
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allowed by Code Case 1605 and (b), not specifying a maximum ultimate tensile strength and relyjng on the bolting material's low alloy steel chemistry, heat treatment and toughness requirements to control f
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ultimate tensile strength ar(acceptable to the staff. ' Compliance with these recommendations ~and the' applicants' alternative approaches satisfythequalitystandardsrequirementsofGDC1,GDC30,and 650.55a; the prevention of fracture of the RCPB requirement of GDC 31; and the requirem'ents of ppendix G,10 CFR Part 50, as detailed in the provisions of the ASME Code, Sections II and III.
5.4.2.1 Steam Generator Materials
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The staff concludes that the steam generator materials specified are acceptable and meet the requirements of GDC 1, 14, 15, and 31, and Appendix B to 10 CFR Part 50.
This conclusion is based on the following:
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The applicant has met the requirements of GDC 1 with' respect to codes and standards by assuring that the materials selected for l
use in Class 1 and Class 2 components were fab.ricated and inspected in conformance with codes, standards, and specifica-s tions acceptable to the staff.
Welding qualification, fabrication, and inspection during manufacture and assembly
- of the steam generators were done in conformance with the
.;requ remen s of Section III and IX of the ASME Code.
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The requirement's of GDC 14 and 15 have been met to assure that the reactor coolant boundary and associated auxiliary systems hav.e been designed, fabricated, erected, and te'sted'so as~ to have an extremely low probability of' abnormal leakage of rapid failure and of gross rupture, during normal operation and anticipated operational occurrences.
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The primary side of'.the steam generator is designed and fabricated to comply with ASME Class 1 criteria as required by the staff.
The secondary side pressure boundary parts of the steam generator are designed, manufactured,'and tested to ASME Code Class 2.
The crevice between the.tubesheet and the inserted tube is minimal because the tube was expanded to the full depth of insertion of the tube in the tubesheet.
The tube expansion and subsequent positive contact pressure between the tube and the tubesheet preclude a buildup of impurities from forming
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in the crevice region and reduce the probability of crevice boiling. The tubes are seal welded to the tube sheet to assure maintenance of separate paths between the primary and secondary water flow.
'The tube support plates will be manufactured from ferritic
-stainless material which has been shown in laboratory tests
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The tube support plates will be designed and manufactured
' with broached holes rat'her than drilled holes. 'The ' broached
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g-hole design' promotes high, velocit9 flow along the bbag -
sweeping impurities \\away from the support plate locations.
3.
The requirements of GDC 31 have been met with respect to the
-fracture toughness of the ferritic materials since,the pressure boundary materials of ASME Class 1 components of the steam generators will comply with the fracture toughness requirements and tests of Subarticle NB-2300 of Section III of the Code.
The materials of the ASME Class 2 components o'f the, steam l
generators will comply with the fracture toughness requirements of Subarticle NC-2300 of Section III of the Code.
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The requirements of Appendix B of 10 CFR Part 50 hav'e been met
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since the onsite cleaning and cleanliriess controls during fabri-cation conform to the recommendations of Regulatory Guide 1.37,
" Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants"
-or the applicant's alternative approaches are acceptable to
,.'.-the staff as discussed in 4.5.1.
The controls placed on the
.. secondary coolant chemistry are in agreement with staff tech-nical positions.
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Reasonable assurance of the satisfactory performance of the steam generator tubing and other generator materials is provided by the
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des.ign' provisions and the manufacturing requiremt nts of the ASME Code and rigorous secondary water monitoring and control < The controls described above combined with conformance with applicable
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codes, standards, staff positions, and regulatory guides constitute an acceptable basis for meeting in part the requirements of General DesignCriter,ia1,14,15,and31,andAppendixB,10CFR'Part50.
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-6.1.1 Engineered Safet'y Features The staff concludes that the engineered safety features, materials specified are acceptable and meet the requirerents of GDC 1, 4, 14, 31, 35, and 41 of Appendix A of 10 CFR Part 50; Appendix B of 10 CFR Part 50, and 10 CFR Part 50, 50.55a.
This conclusion is based on the following:
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General Design Criteria 1, 14, and 31, and 10 CFR Part 50, 50.55a
.have been met with respect to assuring an extremely low probability of leakage, of rapidly propagating failure, afd of gross rupture.
The materials selected for engineered safety features satisfy Appendix-I
,, d f Section III and Parts A, B, and C of Section II of the ASME Code or ccequivalent American Society of Testing and Materials specifications.
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The applicant-has complied with the staff position th'at the yiel'd strength of cold-worked stainless steels shall be less than 90,000 psi.
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In this time frame,'the Code allowed waivin'g of impact testing of Class 2 and 3.
However, based upon the results of impact
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testing by other applicants of thelsame specifi~ cation ~ steels,
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i and correlations of the'metallurg*ical characterization _sf
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these steels with the fracture toughness data presented in NUREG-0577, we conclude that the fracture toughness properties of the ferritic materials in the engineered safety features
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have adequ5te margins against the possibility of nonductile behavior and rapidly propagating. fracture.
The controls on the use and fabrication of the austenitic stainless steel of the. systems satisfy most of the recommen-
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dations of Regulatory Guides 1.31, " Control of Ferrite Content l
in Stainless Weld Metal" and 1.44, " Control of the Use of Sensitized Stainless Ste'el " The alternative approaches taken by the applicant have been reviewed and are acceptable to the staff (see 4.5.1).
Fabrication and heat treatmen't practices performed provide ' ass'urance that the probability of stress corrosio'n cracking will be reduced during the postulated accident time interval.
..Conformance with the Codes and Regulatory Guides and with
, g d.he staff. positions mentioned above, constitute an acceptable
_ sbasis for meeting the requirements of General Design Criteria o
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19 1, 4, 14, 35, 4'1; Appendix B to 10 CFR part 50, and 10 CFR Part 50, $50.55a, in which the systems are to be designed,
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fabricited,'and erected so that the systems can perform t' heir functions as required.
2.
General Design Criteria 1, 14, 31, and Appendix B to 10 CFR Part 50 have been met with respect to assuring that the m attor 1
coolant pressure boundary and associated auxiliary systems have an extremely low probability of ledicage, of rapidly propagating failure, and of gross rupture.
The controls placed on concen-trations of leachable impurities in nonmetallic thermal insulation, used on components of the engineered safety features are in accord-ance with the recommendations of Regulatory Guide 1.36,
" Nonmetallic Thermal Insulation for Austenitic Stainless Steels" or the applicant's alternative approaches are acceptable to the staff as discussed in 5.2.3.
Compliance with the recommendations of Regulatory Guide 1.36 forms a basis for meeting the requirements of GDC 1,14, and 31.
Protective coating systems are discussed in 6.1.2.
3.
.The requirements of GDC 4, 35, 41 and Appendix B to 10 CFR Part 50 viri
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' environmental conditions associated with normal operation, maintenance, testing, and postulatedtaccidents, " incl'uding' loss'-of-coolant accidents.
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'20 The controls on the pH and chemistry of the" reactor containment sprays candtheemergencycorecoolingwater'followingaloss.of-coola[itor
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design' basis accident are adequate. to r, educe the probability of stress
. s 's corrosion cracking of adstenitic (stainless steel components and welds
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of the engineered safety f'eatures systems in containment throughout
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the duration of the postulated accident to completion of cleanup.
Also, the c'ontrols of the pH of the sprays and cooling water,' in conjunction with controls on selection of containment materials, are in accordance with Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident,"
and provide assurance that the sprays and cooling water will not give
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rise to excessive hydrogen ga' evolution resulting from corrosion of s
containment metal or cause serious deterioration of the materials in containment.
The controls placed upon component and system cleani'ng are in accord-ance with the recomme'dations of Reguiatory Guide 1.37, " Quality n
Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants" or the applicant's 1
alternative approaches have been reviewed and approved by the staff
,as discussed in 4.5.1.
These controls provide a basis for the.
Inding that the components and systems have been protected against i
-damage or deterioration by contaminants as stated in the cleaning requirements of Appendix B, 10 CFR Part 50.
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10.3.6. Main Steam a'nd Fee'dwater Materials The staf.f concludes that the main steam and.feedwater system materials
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areacceptableandmeettherelevantreqEirementsof10'CFRJart50, 50.55a,GeneralDesignCriteria1,andAppendixBto10CF[Part50.
This conclusion is based on the following:
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Theapplicantseiectedma'teriaisforClass2and3componentsofthe
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steamandfeedwatersystemsthatsatisff.'AppendixIofSectionIII of the ASME Boiler and Pressure Vessel Code, and meet the require-mentsofPartsA,B,'andCofSectonI5oftheCode.
Conformance to the recommendations of Regulatory Guide 1.85 is discussed in 5.2.1.2.
In this time frame, the Code allowed waiving of impact testing of main steam and feedwater materials.
However, based upon the results of impact testing by other applicants of the same specification steels,andcorrelationsofthemetallurgicalcharacterizationof these steels with the fracture toughness data presented in NUREG-0577, we conclude that the fracture toughness properties of the ferritic materials in the main steam and feedwater systems which were no't
. impact,_ tested have adequate safety margins'against the possibility o,f nonductile behavior and rapidly propagating fracture.
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The applicant has satisfied to the extent practiEal, the recommend-ations of Regulatory Gui,de 1.71, " Welder Qualification for Areas of Limited Accessibility" by meeting the regulatory positions in V, ',
- t Regulatory Guide'1.71 or by 'mie,eting alhe;rnative approaches.which the staff has reviewed.a'qd found to be acceptable (see 4.5.1).
The onsite cleaning and cleanliness controls during fabrication satisfy the positions given in Regulatory Guide 1.37, " Quality
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Assurance Requir'ements for Cleaning of Fluid Systeas and Associated
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Components of Water-Cooled Nuclear Power 'P1 ants," and the requirements of ANSI Standard N 45.2.1-1973, " Cleaning of Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants" or the applicant ~ alternative approaches have been reviewed and are acceptable to the staff as discussed on 4.5.1.
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