ML19308E102

Forwards Materials Engineering Branch Review of SER & Tech Specs.Ser Complete;Tech Specs Incomplete

ML19308E102
Person / Time
Site:	Crystal River
Issue date:	02/15/1974
From:	Maccary R US ATOMIC ENERGY COMMISSION (AEC)
To:	Moore V US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8003200825
Download: ML19308E102 (15)

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L~~t 2-3; L "uchley Foquested Corpi.etion Data-iebruary 15, 1974' Description of Tu h S.:Cuty t. valuation &.Tochnical :ipecifications Naview ReviW Statun:

SEE Cer*.plete: Te.chnical Spe.cifications In:c-plete-Tim infora.?-% cubmittc 4 hy the applicant, including.budment ::o. 36, has be re ev red Lv the Fe:crialr. C 71necring French, Directorate of.

Licenstne.

Dr coettens of th. Mty Preluation are enclocM.

The arpitenet has net yet iubmitted a ecc:2. te met of Scimien1 Specifica-tiene. The enciesed c.ycents are on the dr:J Technical Sp e ifications

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DOCKET No. 50-302 SAFETY EVALUATION MATERIALS ENGINEERING BRANCH, L REACTOR Reactor Vessel Internals General Material Considerations We have reviewed the selection of materials for the reactor vessel internals required for reactor shutdown and components relied upon for adequate core cooling. All materials are compatible with the reactor coolant, and have performed satisfactorily in similar applications.

Undue susceptibility to intergranular stress corrosion cracking has been prevented by avoiding the use of sensitized stainless steel according to y...f the 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 experi-ence and avoidance of sensitization by the methods recommended in Regulatory Guide 1.44, provides reasonable assurance that the reactor vessel internals will not be susceptible to failure by corrosion or stress corrosion cracking.

The applicant has described the measures that were taken to ensure that deleterious hot cracking of austenitic steel welds was prevented. All weld filler metal was of selected composition, and welding processes were controlled to produce welds with adequate delta ferrite, in conformance u

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,n with the recorrnendation in Regulatory Guide 'l.31, " Control of Stainless

. Steel Nelding." Following these recon:mendations provides reasonable

. assurance that no deleterious hot cracking will be present that could contribute to loss of integrity or function' l capability.

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REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS Integrity of Reactor Coolant Pressure Boundary Fracture Toughness 1.

Compliance with Code Requirements We have reviewed the materials selection, toughness requirements, and extent of materials testing accomplished by the applicant to provide assurance that the ferritic materials used for pressure retaining components of the reactor coolant boundary will have adequate toughness under test, normal operation, and transient conditions. All ferritic materials, not including piping, were ordered and tested in accordance with the requirements of the ASME Soiler and Pressure Vessel Code,Section III (1965 Edition and with Addenda through Summer 1967).

Piping met the requirements of ASAS Standard B31.7, dated February 1968, including the Errata dated June 1968. Dropweight NDL data were obtained for the beltline shell plates of the reactor vessel.

The fracture toughness tests and procedures required by Section III of the ASME Code, augmented by the additional dropweight testing for the reactor vessel, provide reasonable assurances that adequate safety margins against the possibility of nonductile behavior or rapidly propagating fracture can be established for the pressure-retaining co=ponents of the reactor coolant boundary.

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Operating Limitations The res.ctor will be operated in a manner that will minimize the possibility of rapidly propagating failure, in accordance with

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4-Appendix G'to Section III of the ASME Boiler and Pressure Vessel Code, Summer 1972 Addenda, and Appendix G, 10 CFR 50. Additional conservatism in the pressure-temperature limits used for heatup, cooldown, testing, and core operation will be provided because these will be determined assuming that the beltline region of the reactor vessel has already been irradiated.

The use of Appendix G of the Code'as a guide.in establishing safe operating limitations, using results of the fracture toughness tests performed in accordance with the Code and AEC Regulations, will ensure adequate safety margins during operating, testing,' maintenance, and postulated accident conditions. Compliance with these Code provisions i

and AEC regulations, constitute an acceptable basis for satisfying the requirements of AEC General Design Criterion 31, Appendix A of 10 CFR Part 50.

3.

Reactor Vessel Material Surveillance Program The toughness properties of the reactor vessel beltline material will i

be monitored throughout service life with a material surveillance program-that will comply with the intent of Appendix H,10 CFR 50 (July 17,1973).

The program is consistent with other surveillance programs that have i

been found acceptable for other PWR plants.

The copper content of the reactor vessel beltline has been determined, but the number of capsules provided in the surveillance program is conservatively based on assuming high values of sensitivity.

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, ' Changes in_the fracture toughness of material in the reactor vessel bel'tline caused by exposure to neutron radiation will be assessed properly,-and adequate '- safety margins against the possibility of vessel failure will be provided sinc.e the essential material surveil-lance requirements of Appendix H, 10 CFR Part 50, are met.

The surveillance progran constitutes an acceptable basis for monitoring.

radiation induced changes in the fracture toughness of the reactor vessel material, and will satisfy the requirements of AEC General Design Criterion 31, Appendix A,' of 10 CFR Part 50.

Although the use of naterial of known copper content for the reactor vessel beltline will minimize the poscibility that rhdiation will cause sericus degradation of the tougbaess properties, the applicant has stated that should results of tests indicate that tb+ toughness is not-adequate, the reactor vessel can be annealed to restore the toughness to acceptable levels.

General Material Considerations We have reviewed the materials of construction for the reactor coolant pressure boundary' to ensure that the possibility of serious corrosion or stress corrosion is minimized. All materials used are compatible with the expected environment, 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

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' because sensitization will be avoided, and adequate precautions will be-taken to prevent contamination during manufacture, shipping, storage, and cons truction.

The plans to avoid sensitization are in general conformance with Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless.

Steel," and include controls on compositions, ht; treatments, welding processes, and cooling rates.

The use of materials with satisfactory service experience, and the high degree of conformance with Regulatory Guide 1.44, " Control of Sensitized Stainless Steel," provide reasonable assurance that austenitic stainless steel components will be compacible with the expected service environments, and the probability of loss of structural integrity is minimized.

Water Chemis try Control Further protection against corrosion problems will be provided by control of the chemical environment.

The ccuposition of the reactor coolant will be controlled; and the proposed =aximum contaminant levels, as well as the. proposed pH, hydrogen overpressure, and boric acid concentrations, have been shown by tests and service experience to be adequate to protect against corrosion and stress corrosion problems.

We have evaluated thel proposed requirements for the external insulation used on austenitic stainless steel components.

Chloride and silicate

- contents will be controlled.

The possibility that serious corrosion or stress corrosion problems would occur in the unlikely event that ECCS or containment spray system operation.

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i, is required.wi11 be minimized 'because the-pH of the circulating coolant -

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will be maintained above 9.0 by~ hydroxide additions.

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The applicant has shown that the secondary water chemihtry will be con-trolled to prevent stress corrosion of the steam-generator. tubing, and that the adequacy of the compositional limits used'has been proven by i

satisfactory service' experience.

The controls on chemical composition that vill' be imposed on the reactor coolant, secondary water, emergency core cooling water, and the use of low chloride external. thermal insulation, provide reasonable assurance that the e

reactor coolant boundary materials will be adequately protected frca con-ditions that would lead to loss of integrity from stress corrosion.

Centrol of Stainless Steel Welding We have reviewed the controls proposed to prevent hot cracking (fissuring) of 'austenitic s teel welds.

These precautions include control of weld metal composition and welding processes to ensure adequate delta-ferrite content in the weld metal.

The proposed methods comply with Section III I

of the ASME Code, and are in essential conformance with Regulatory Guide 1.31, " Control of Stainless Steel Welding." The use of materials, processes, and test methods that-are in accordance with these require-ments and recommendations will provide reasonable assurance that loss of

' integrity of austenitic stainless steel welds caused by hot cracking Lduring welding wil1~ not occur.

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The probability of a loss of pump flywheel integrity can be minimized by the use of suitable material, adequate design, and inservice inspection.

The applicant has stated that the integrity of the reactor coolant pump flywheel is provided by having designed for a 125% overspeed condition while the maximum anticipated overspeed is 110% of normal speed.

In the unlikely event of a 125% overspeed condition the maximum primary stress at the bore is approximately 70% of the yield strength, the flywheel was purchased prior to the requirements of Regulatory Guide 1.14 which permits 2/3 of yield at the design overspeed condition.

In addition, a 100%

ultrasonic volumetric inspection of the flywheel, using ASME Section III acceptance criteria, was performed.

Inservice inspections of the flywheel will be performed in accordance with the provisions of Regulatory Guide 1.14.

We conclude that the provisions for material selection and flywheel design, and the 'sse of a Regulatory Guide 1.14 inservice inspection program ensure adequate flywheel integrity.

Reactor Vessel and Appurtenances Reactor Vessel Intecrity We have reviewed all factors contributing to the structural integrity of the reactor vessel and we conclude there are no special considerations that make it necessary to consider potential vessel failure for Crystal River Unit 3.

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The bases for.our conclusion are that the design, material,-fabrication, inspection, and quality assurance requirements will conform to the rules of the ASME Boiler and Pressure Vessel Code,Section III, all addenda thrcugh Su=mer 1972, and all applicable Code Cases.

The fracture toughness requirements of the ASME Code,Section III, 1965 Edition,have been met.

Also, operating limitations on temperature and pressure will be established for this plant in accordance with Appendix G,

" Protection Against Non-Ductile Failure," of the 1972 Su=mer Addenda of the ASME Boiler and Pressure Vessel Code,Section III, and Appendix G, 10 CFR 50.

The integrity of the reactor vessel is assured because the vessel:

1.

Will be designed and fabricated to the high standards of quality required by the ASME Boiler and Pressure Vessel Code and pertinent Code Cases listed above.

2.

Will be made from materials of controlled and demonstrated high quality.

3.

Will be extensively inspected and tested to provide substantial assurance that the vessel will not fail because of material or fabrication deficiencies.

4.

Will be operated under conditions and procedures and with protective devices that provide assurance that the reactor vessel design con-ditions will not be exceeded during normal reactor operation or

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during most upsets in operation, and that the vessel will not fail under the conditions of any of the postulated accidents.

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Will be subjected to monitoring and periodic inspection to demonstrate that the high initial quality of the reactor vessel has not deteri-r orated significantly under the service conditions.

6.

May be annealed to restore the material toughness properties if this becomes necessary.

Inservice Inspection Program To ensure that no deleterious defects develop during service, all welds will be inspected periodically.

The applicant has stated that the design of the reactor coolant system incorporates provisions for access for inservice inspections in accordance with Section XI of the'ASME Boiler and Pressure Vessel Code, and that methods will be provided to facilitate the remote inspection of those creas of the reactor vessel not readily accessible to inspection personnel. The conduct of periodic inspections and hydrostatic testing of pressure retaining components in the reactor coolant pressure boundary in accordance with the require =ents of ASME Section XI Code provides reasonable assurance that evidence of structural degradation or loss of leaktight-integrity occurring during service.will be 1

detected in time to permit corrective action before the safety function of j

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a component is compromised..Coupliance with the inservice inspections required by this Code constitutes an acceptable basis for satisfying the

. requirements of AEC General Design Criterion 32, Appendix A of 10 CFR I

l Part 50.

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RCPB Leakage Detection System Coolant leakage within the containment may be an indication of a small through-wall flaw in the reactor coolant pressure boundary.

The leakage detection system proposed for intersystem leakage is by means of radioactivity monitors and flow and level monitors.

The leakage system proposed for leakage to the containment will include diverse leak detection methods, will have sufficient sensitivity to measure small leaks, will identify the leakage source to the extent practical, will be provided with suitable control room alarms and readouts.

The major components of the system are the containment airborne particulate and gas radioactivity monitors, and level and flow indication on the containment sump.

Indirect indication of leakage can be obtained frem the contain=ent humidity, pressure, and temperature indicators.

The leakage detection systems will provide reasonable assurance that any structural degradation resulting in leakage during service will be detected in time to permit corrective action satisfying the requirements of AEC General Design Criterion 30, Appendix A of 10 CFR Part 50, and is thus acceptable.

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' ENGINEERED SAFETY FEATURES Containment Design Evaluation Containment Leakage Testing Program The containment design includes the provisions and features planned which satisfy the testing requirements of Appendix J, 10 CFR Part 50.

The design of the containment penetrations and isolation valves permits individual periodic leakage rate testing at the pressure specified in Appendix J, 10 CFR Part 50.

Included are those penetrations that have resilient seals and expansion bellows, i.e., air locks, emergency hatches, refueling tube blind flanges, hot process line penetrations, and electrical penetrations.

The proposed reactor centainment leakage testing program complies with the requirements of Appendix J to 10 CFR Part 50.

Such compliance provides adequate assurance that containment leaktight integrity can be verified 1

throughout service lifetime and that the leakage rates will be periodically checked during service on a timely basis to maintain such leakages within the specified limits.

Maintaining containment leakage rates within such limits provides reasonable assurance that, in the event of any radioactivity releases within the con-tainment, the loss of the containment atmosphere through leak paths will not be in excess of acceptable limits specified for the site.

Compliance with the requirements of Appendix J constitutes an acceptable basis for-satisfying the requirements of AEC General Design Criteria 52, 53, and 54, Appendix A of 10 CFR Part 50.

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CONTAIMIENT HEAT REMOVAL AND ECCS SYSTDIS Ceneral' Material Considerstic-= ',

(Compatibility with coolari We have reviewed the matettals selection proposed for the containment heat removal and ECCS systems, in conjunction with the expected chemistry of the cooling and contaitacut opcay system water. The applicant has shown that the use of sensitized scainless steel will be avoided, and that the pH of the containment spray and the circulating coolant will be con-trolled by sodium hydroxide additions.

There are test data verifying that the preposed chemistry will not cause stress corrosion cracking of austenitic stainless steel un a-acnditions that would be present during accident d

conditions.

I We have concluded that the controls on material and cooling water chemistry-proposed will provide 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 stated F; '.'veLIIng of austenitic stainless steel for-components of these systesv will be controlled to prevent deleterious hot cracking. The proposed control of weld metal composition and welding procedures are in general confor=ance with the reco=mendations of Regulatory Guide 1.31, " Control of Stainless Steel Welding," and will provide assurance that lost

.c function will not result from hot cracking of welds.-

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TECHNICAL SPECIFICATIONS, 15.3.1.3 Pressurization. Heatup, and Cooldown Limitations This Technical Specification is incomplete. We are awaiting heatup and cooldown curves drawn to the requirements of 10 CFR 50, Appendix G.

The applicant has promised to submit the curves by mid April 1974 and in Amendment No. 36 to the FSAR stated, " Technical Specification 15.3.1.3 will be revised to be in accord with Appendix G of 10 CFR 50 as practicable."

15.3.1.4 Leakage 15.3.1.5 Chemis try 15.4.4.1 Reactor Coolant System Inteerity Testing - ISI 15.4.4.2 Testing Following Opening of System The four above listed Technical Specifications are acceptabl.i.

15.4.2.2 containment Leakace Tests The exponent on the equation in A.2.c ss 1/2 not 1/3.

The ie=ainder of this Technical Specification is acceptable.

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