ML19319D528

Forwards Request for Addl Info for FSAR Review Through Amend 26.Previous Response Inadequate to Questions 4.34 & 6.1 Re pressure-temp Limitations & Engineered Safety Features Inservice Insp,Respectively.Response Required by 730807

ML19319D528
Person / Time
Site:	Crystal River
Issue date:	06/08/1973
From:	Maccary R US ATOMIC ENERGY COMMISSION (AEC)
To:	Deyoung R US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8003170654
Download: ML19319D528 (7)

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T Richarl C. %'Toung, Assistant Director-for Wec.,arized Water Reactors Directorate of Licensing FLORIDA PO W R CORPORATION - CRTSTAL RIVER UNIT 3 (OL), W KET No. 50-302 m'

Plant Masse Crystal River Unit 3 s

Licensing Stege: OL Docket Number:

50-302 Responsible Eranch and Project Manager: P W-4; B. C. Buckley Requested Co.:pletion Date: June 8, 1973 Description of Reaponse: Questions - Second Round (Q2)

Review Status: Awaiting Information Infor ation submitted by the applicant in the FSAR, including Amendment -

No. 24, h.u bv.n revie.ed. by the Materi:1s Enginering Ernneh, Directornt-of Licensing..Mcquate r~spcnsas to the enclosed request for additional infor::2ation.uc. required before we can complete c : enInstion. In order to maintain the current schelule, the apt,11 cant should respond to this request by August 7, 1973.

The applicant in response to ousation 4.34 regarding pro paed pres:. m*-

temperatur2 limitati e durin6 reactor startup, shutdown, and hydro h cic testing stated " Appendix G of the iumner 1972 Addenda.Section III of ASME Code, was not used as a guide'...."

Thic recponr.a is unacceptable.

The operating and hydrcatatic test limitations must be at least as con-servative as th)se developed using Appendix G of the Coda'as a guide.

.The applicant's response to question 6.t resarding tha inservice inspac-b tion.of engineered safety features is' unsatisfactory because ha did not.;.1 agree to perform these: inspections to the dag::ee of acca!Esnbility designeds,

.into the plant, and.as required;by ASME Code seetion IIe 1972 Sinter gi 'a T

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CRYSTAL RIVER NUCLEAR GENERATING PLA'iT, USlI 3 (OL)

DOCKET No. 50-302-SECOND REQUEST FOR ADDITIONAL INFORMATION (Q2)

MATERIALS ENGINEERING BRANCH, L REACTOR COOLANT SYSTE'! AND CONECTED SYSTE"S Ceneral Material Considerations 1.

Material Snecifications In addition to the specifications already listed in the FSAR for principal pressure retaining ferritic =aterials and austenitic stainless steels used for components that are part of the reactor

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coolant pressure boundary, provide RCPB specifications for the wcld materials for fabrication and asseably of the components.

L'ith respeat to ferritic materials (including welds) of the reactor pressure vessel beltline, provide information regardin?, the speci-fications for these caterials showing additionally imposed limits on residual elements (reportable and nonreportable) by specification requirements which are intended to reduce sensitivity to irradiatien embrittlement in service Any additional or special requirements by the purchaser should also be indicated.

2.

Compatibility with External Insulation and Environmental Atmosphere Provide a description of the compatibility of the reactor coolant pressure boundary naterials of construction with the external insulation or the environmental at=osphere in the event of coolant leakage.

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,, REACTOR COOLANT SYSTDI AND COR;ECTED SYSTDiS Austenitic Stainless Steel For all a'ustenitic stainless steel used for components that are part of (1) the reactor coolant pressure boundary, (2) systems required for reactor shutdown, (3) systems required for emergency core cooling, (4) reactor vessel internals required for emergency core coolinF, and (5) reactor vessel internals relied on to permit adequatu.-: cooling for any acde of normal operation or under postulated accident conditions, the following information should be provided:

1.

Cleaning and Contamination Prctection Procedures Describe the procedures that were used and will be used to ensure that the material was and will be suitably cleaned and protected against contaminants capable of causing stress corrosion cracking throughout the fabrication, shipment, storage, construction, testing, and operation of co=ponents and syste=s.

- 2.

Avoidance of Sensitization Provide a description of materials, processes, inspections, and tests

.that were used to ensure freedom from the increased susceptibility to intergranular stress corrosion caused by sensitization.

This should include the-following:

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If special processing or fabrication methods were used that subjeeted the material to temperatur"es between 800 and 1500*F, or that involved slow cooling from temperatures over 1500*F, provide justification that such treatments did not cause increased susceptibility to intergranular stress corrosion.

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Indicate special requirements on chemical analysis for any materials that during normal operation will be exposed to water environments containing over 0.10 ppm dissolved oxyg when at temperatures over 200*F.

If the presence of delta ferrite was relied on to prevent c.

sensitization of welds or castings, describe the methods that were used to ensure the presence of at least 5% delta ferrite.

3.

Weldine of Austenitic Stainless Steel Describe the procedures and requirements that were employed to avoid

, hot cracking of austenitic stainless steel velds, especially pertain-ing to filler metal compositions, welding procedure qualifications, and methods for ensuring adequate delta ferrite content of production welds.

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4-REACTOR COCL:T SYSC! A'O Cor'ECTD SYSTD!S Inservice Insocction Procram In response to Question 6.1 in Amendment No. 24, the statement appears that, "A formal inservice inspection program for fluid systems othcr than those composing the reactor coolant pressure boundary system is not planned...,"

This is not acceptable. Our position is that these inspections should be performed and follow the AS!E Section XI,1972 Winter Addenda, " Rules for Inservice Inspection of Nuclear Power Plant Components," within the limits of accessibility designed into the Crystal River, Unit 3 nuclear power plant.

Provide docu=entation that Crystal River 3 will comply with the above Regulatory positien and will include the 1.'.spections in the Technical Specifications.

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REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS Fracture Touchness Inresponse.thQuestion4.34inAmendmentNo. 26 the.following statement is made:

" Appendix G of the Summer 1972 Addenda Section III of ASME Code, was.not used as a guide in establishing the operating pressure and temperatur'e limitations of the reactor coolant systen." No operating curves were in Amendment No. 26, but the text indicated that the pressure-temperature limitations were less conservative than what Appendix G would have given. Hydrostatic and leak testing limitations were not centioned.

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. Pressure-temperature operational limitations during reactor startup and

- shutdcwn, and during hydrostat1 and leak testing of the reactor ecolant l

syster,must be at least as conservative as those determined in conformance I

with Appendix G, " Protection Against Nonductile Failure," of Section III of the ASME Boiler and Pressure Vessel Code, Surner 197.2 Addenda.

Provide doeurentation that the above Regulatory Position will be adhered to and wili Se included in the Technical Specifications for operation of Crystal 4

River 3.

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