ML20210T334

Forwards Matls Performance Section of Matls Engineering Branch Section of Safety Evaluation Re PSAR Through Amend 14.Integrity of Reactor Vessel Assured for Listed Reasons

ML20210T334
Person / Time
Site:	Satsop
Issue date:	05/15/1975
From:	Maccary R Office of Nuclear Reactor Regulation
To:	Deyoung R Office of Nuclear Reactor Regulation
References
CON-WNP-1775 NUDOCS 8605300064
Download: ML20210T334 (12)

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Information ===hmteted by the applicant in the PSAR through AmenAmong '

No. 14 has been reviamed by the MatartmIn Performance Section'af the

" Materials Engineering Branch,"Offies of Muclear Reactor Regulation. 'f Our sections of the' Safety Evaluation are snelosed. ,

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l COMBUSTION ENGINEERING 6

WPPSS NUCLEAR PROJECTS NUMBERS 3 AND 5 (CP)

DOCKET NUMBERS 50-508/509 i SAFETY EVALUATION i 4 MATERIALS ENGINEERING BRANCH MATERIALS PERFORMANCE SECTION REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

~i l Integrity of Reactor Coolant Pressure Boundary i o

i Fracture Toughness l

i 1. Compliance with Code Requirements l i We have reviewed the materials selection, toughness requirements, i  !

and extent of materials testing proposed by the applicant to provide  ;

assurance that the ferritic materials used for pressure retaining ,

components of the reactor coolant boundary will have adequate  ;

i toughness under test, normal operation, and transient conditions.  ;

The ferritic materials are specified to meet the toughness require- l ments of the ASME Code,Section III, including Summer 1972 Addenda.

In addition, materials for the reactor vessel are specified to meet e 'l the additional test requirements and acceptance criteria of Appendix G, 10 CFR 50. ,

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! The fracture toughness tests and procedures required by Section III

' a l of the ASME Code, as augmented by Appendix G, 10 CFR 50, for.the

reactor vessel, provide reasonable assurances that adequate safety

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.' margins against the possibility of nonductile behavior or rapidly i

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. propagating fracture can be established for all pressure retaining I

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• components of the reactor coolant boundary. '

2. Operating Limitations ,

I The reactor will be operated in accordance with the ASME Code,Section III, including Sussaar 1972 Addenda, and Appendix G, >

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l 10 CFR 50. This will minimize the possibility of failure due to e 4

rapidly propagating crack. Additional conservatism in the pressure- ,

temperature limits used for heatup, cooldown, testing, and core operation will be provided because the pressure-temperature limits 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 NRC Regulations, will i ensure adequate safety margins during operating, testing, maintenance, l

and postulated accident conditions. Compliance with these Code I

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

3. Reactor Vessel Material Surveillance Program The toughness properties of the reactor vessel beltline material will be monitored throughout service life with a material r .

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.j surveillance program that will meet the requirements of ASTM E 185-73 '

and comply with Appendix H, 10 CFR 50.

, l Changes in the fracture toughness of material in the reactor vessel beltline caused by exposure to neutron radiation will be assessed properly, and adequate safety margins against the possibility of vessel failure can be provided if the material requirements of the

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i above documents are met. Compliance with these documents will i

ensure that the surveillance program constitutes an acceptable basis for monitoring radiation induced changes in the fracture toughness ,

of the reactor vessel material, and will satisfy the requirements of .

NRC General Design Criterion 31, Appendix A of 10 CFR Part 50.

Although the use of controlled composition material for the reactor vessel beltline will minimize the possibility that radiation will cause serious degradation of the toughness properties, the .

applicant has stated that,should results of tests indicate that the toughness is not adequate, the reactor vessel can be annealed to I restore the toughness to acceptable levels.

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4. Reactor Vessel Closure Studs ,

The reactor vessel closure studs will be procured and inspected in  ;

conformance with the mechanical and toughness propert.y requirements  ;

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and inspection requirements recomnended in Regulatory Guide 1.65, i

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" Materials and Inspections for Reactor Vassel Closure Studs."

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5. Inservice Inspection Program To ensure that no deleterious defects develop during service, selected welds and veld heat-affected zones will. be inspected periodically. The applicant has stated that the design of the

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reactor coolant system incorporates proeisions for direct or remote I

sccess for inservice irtspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Coda, and that suitable equipment I

will be Ceveloped to facilitate the remote inspection of those areas of the regctor vessel not readily accessibiu to inspection personnel.

The conduct of periodic inspections and hydrostatic gesting of pres-sure retaining components in the resctor coolant pressure boundary in accordanck with the requiremects of ASME Section XI Code provides reasonable assurance that evidence of structural degradation or loss of leaktig'nt-integrity or. curring during servise will be detected in tihe to permit corrective action before the safety function of a i

,' component is compromiaed. Compliance with the insarvice ,1uspections I required by this Code constitutas an acceptable basis for satisfying i

the requirements of NRC Ceneral Design Criterion 32, Appendix A of 10 CFR Part $0.

To ensure that,no deleterious defects develop during servica .

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. ASME Code Class 2 cysteE components, selected wolds and weld heat-affected zones are inspected prior to reattor startup and periodically

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l throughout the life of the plant. Code Class 2 systems and Code Class 3 systems receive visual inspections while the systems are pressurized in order to detect leakage, signs of mechanical or structural distress, and corrosion. ,

l Examples of Code Class 2 systems are: (1) Rasidual heat removal i

! systems, (2) Portions of chemical and volume control systems, and (3) Engineered safety features not part of Code Class 1 systems.

Examples of Code Class'3 systems are: (1) ' Component cooling water systems and (2) Portions of radwaste systems. All of these systems transport fluids. The applicant has stated that the Code Class 2 systems meet the requirements of ASME Section II. The Code Class 2 systems and Code Class 3 systems are in conformance with the recom-mandations of Regulatory Guide 1.51. Compliance with the inservice inspections required by this Code and Regulatory Guida constitutes an acceptable bases for satisfying NRC General Design Criterion 36,

39, 42, and 45, Appendix A of 10 CFR Part 50.

} 6. Ptsap Flywheel 4

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

inservice inspection.

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The integrity of the reactor coolant pump flywheel is provided by i

compliance with the intent of NRC Regulatory Guide 1.14. " Reactor 6

Coolant Pump Flywheel Integrity."

The use of suitable material, and adequate des 16 n and inservice inspection for the flywheels of reactor coolant pump motors as

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' specified in the SAR provides reasonable assurance (a) that the J

I structural integrity of the flywheels is adequate to withstand the forces imposed in the event of pump design overspeed transient with-4 out loss of function, and (b) that their integrity will be verified periodically in service to assure that tha required level of sound-ness of the flywheel material is adequate to preclude failure.

Compliance with the recommendations of NRC Regulatory Guide 1.14 constitutes an acceptable basis for satisfying the requirements of NRC General Design Criterion 4, Appendix A of 10 CFR Part 50. ,

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

The leakage detection system proposed for leakage to the containment

> vill include diverse leak detection methods, will have sufficient sensitivity to measure small leaks, will identify the leakage source to the extent practical, and will be provided with suitable control .

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! room alarms and readouts. The leakage detection systems proposed to i

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detect leakage from components and piping of the reactor coolant I pressu'a houndary follow the recommendations of NRC Regulatory Guide 1.45 as far as practical and thus provide reasonable assurance that any structural degradation resulting in leakage during service will be detected in time to permit corrective actions. This 1

e constitutes an acceptable basis for satisfying the requirements of NRC General Design Criterion 30, Appendix A of 10 CFR Part 50.

8. Reactor Vessel Appurtenances Reactor Vessel Integrity:

We have reviewed all factors contributing to the structural integrity of the reactor vessel, and we conclude there are no special considera-tions that make it necessary to consider potential vessel failure for WPPSS-Nuclear Projects 3 and 5.

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The bases for our conclusion are that the design, material, fabrication, inspection, and quality assurance requirements for the reactor vessels of WPPSS-Nuclear Projects 3 and 5 will conform to the rules' of the

, 4 f ASME Code,Section III, 1971 Edition, and the 1972 Summer Addenda.

! Also, operating limitations on temperature and pressure will be i l eatablished for this plant in accordance with Appendix G, " Protection AgainstNon-DuctileFailure,"ofthe1972SummerAddendaofjtheASME t

. Boiler and Pressure Vessel Code,Section III, and Appendix G, 10 CFR 50.

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I The integrity of the reactor vessel is assured because the vessel: i l

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a. Will be designed and fabricated to the high standards of quality I required by the ASME Code Section III and pertinent Code Cases listad above.

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

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

d. Will be operated under conditions and procedures and with protective devices that provide assurance that the reactor vessel design conditions will not be exceeded during normal reactor operation or tluring most upsets in operation, and that the vessel will not fail under the conditions of any of the postulated accidents.

a. Will be subjected to monitoring and periodic inspection to r l l

demonstrate that the high initial quality of the reactor vessel

- has not deteriorated significantly under the service conditions.

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

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MATERIALS PERFORMANCE SECTION, MATERIALS ENGINEERING BRANCH

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REFERENCES General Federal Register 10 CFR Part 50, Appendix A, " General Design Criteria for Nuclear Plants," July 7, 1971.

_ 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," Rev. 1, October 1972.

Fracture Toughness 10 CFR 50 - Appendix G, " Fracture Toughness Requirements," June 1, 1973.

ASME Boiler and Pressure Vessef Code,Section III, 1972 Summer Addenda, including Appendix G, " Protection Against Non-Ductile Failure."

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l ASME Specification, SA-370-71b, " Methods and Definitions for Mechanical i

Testing of Steel Products," ASME Boiler and Pressure Vessel Code,

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Section II, Part A - Ferrous, 1971 Edition, Summer and Winter, 1972 f

! Addenda.

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ASTM Specification E 208-69, " Standard Method for Conducting Dropweight

j. Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," Annual Book of ASTM Standards, Part 31, July 1973.

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' ASTM Specification E 23-72, " Notched Bar Impact Testing of Metallic

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.Mderials,"Anan=1 Book of ASTM Standards, Part 31, July 1973.

Material Surveillance. Programs 10 CFR 50 - Appendix H, " Reactor Vessel Material Surveillance Program i Requirements," June 1.-1973.

ASTM ' specification E 185-73, " Surveillance Tests on Structural Materials in Nuclear Reactors," Annual Book of ASTM Standards, Part 30, July 1973.

Pump Flywheels NRC Regulatory Guida 1.14, " Reactor Coolant Pump Flywheel Integrity,"

October 27, 1971.

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b. 'j f RCPB Leakage Detection Systems i i

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NRC Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage i s Detection Systems," May 1973.

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Ins $rviceInspectionProgram NRC Guideline Document, " Inservice Inspection Requirements for Nuclear

}- Power Plants Constructed with Limited Accessibility for Inservice

! Inspect 1'ons," January 31, 1969.

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! ASME Boiler and Pressure Vessel Code,Section XI, 1971 Edition, including 1

1 Winter 1971, Sunsner 1971, Winter 1972, and Summer 1973 Addenda.

J Regulatory Guide 1.51, " Inservice Inspection of ASME, Class 2 and 3 Nuclear Power Plant Components," May 1973.

Reactor Vessel Integrity i ASME Boiler and Pressure Vessel Code,Section III, 1971 Edition plus Addenda through Winter 1972.

ASME Boiler and Pressure Vessel Code,Section XI, 1971 Edition plus Addenda through Winter 1972.

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