ML20210S923
| ML20210S923 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 06/20/1975 |
| From: | Maccary R Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-1703 NUDOCS 8605290512 | |
| Download: ML20210S923 (19) | |
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. COMBUSTION ENGINEERING w
WPPSS. NUCLEAR PROJECTS NUMBERS 3 AND 5 (CP)
DOCKET NUllBERS 50-508/509 x
SAFETY EVALUATION
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- MATERIALS ENGINEERING BRANCH r
MATERIALS PERFORMANCE SECTION e
REACTOR' COOLANT SYSTb! AND CONNECTED SYSTEMS
~3 r Integrity of Reactor / Coolant Pressure Baundary
- Fracture Toughness
.1.
Compliance with Code Requirements We have reviewed the materials s61ection, toughness requirements, and ei.~ent of materials testing proposed by the applicant to provide
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assurance that the ferritic materials used for pressure retaining
<cohponents of the reactor coolant boundary will have ~ adequate tough-nessuYidertest, normal operation, and transient conditions. The s
ferritic materials are specified to meet the toughness requirements of the ASIC Code,Section III, including Summer 1972 Addedna. In addition, materials for the reactor vessel are specified to meet.
th'e'idditional test requirements and acceptance criteria of Appendix C,-10 CFR 50.
The, fracture toughness tests and procedures required by Section III of the-ASME Code, as augmented by Appendix G, 10 CFR 50, for the reattor vessel, provide reis'onabic assurances that adequate safety margknsagainstthepossibilityofnonductilebehaviororrapidly 0
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n propagating fracture can be established for all pressure retaining components cf the reactor cool, ant boundary.
2.
Operating' Limitations The safety evaluation for operating limitations is presented in the Safety Evaluation for CESSAR.
3.
Reactor Vessel laterial Surveillance Program The safety evaluation for reactor vessel material surveillance program-is presented in the Safety Evaluation for CESSAR.
4.
Reactor Vessel Closure Studs The safety evaluation for reactor vessel closure studs, is presented in the Safety Evaluation for CESSAR.
Inservice Inspection Program To ensure.that no deleterious defects develop during service, selected welds and weld heat-affected zones wil'1 be inspected periodically. The applicant has stated that the design of the reactor coolant system incorporates provisions for direct or remote access for inservice inspect-ions in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, and that suitabic equipment will be developed to facilitate the remote inspection of those areas of the reactor vessel not readily accessibic 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 requirements of ASME i
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Section XI Code provides reasonabic assurance that evidence of structural degradation or loss of leaktight-integrity occurring during service will l
be detected in time to permit corrective action before the safety function of a component is compromised. Compliance with the inservice inspections required by this Code constitutes an acceptable basis for satisfying the requirements of NRC Cencral Design Criterion 32, Appendix A of 10 CFR Part 50.
To ensure that no deleterious defects develop during service in ASME I
Code Class 2 system components, selected welds and weld heat-affected zones will be inspected prior to reactor startup and periodically throughout the life of the plant. In addition Code. Class 2 systems and i:
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Code Class 3 systems will receive visual inspections while the systems are pressurized in order to detect leakage, signs of mechanical or structural distress, and corrosion.
Exampice of Code Class 2 systems are:
(1) Residual heat removal systems, (2) Portions of chemical and volume control systems, and (3) Enginected i!
safety features not part of Code Class 1 systems. Examples of Code
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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 design of Code Class 2 systems meets the requirements of ASME Section XI.
Also, the Code Class 2 systems and
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Code Class 3 systems.are in conformance with the recommendations of
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Regulatory culde 1.51.
Compliance with the inservice inspections I
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required by this Code and Regulatory Guide constitutes an acceptable
. bases for satisfying NRC.Ceneral Design Criterion 36.-39, 42, and 45, Appendix A'of 10 CFR Part 50.
' Pump Flywheel The safety evaluation for pump flywhel is presented in the Safety Evaluation for CESSAR.
RCPB Leakage Detection System Coolant leakage within the primary containment may be an indication of a small through-wall flaw in the reactor coolant pressure boundary.
1 The-leakage detection system proposed for leakage to the containment will include diverse. leak detection methods, will have sufficient senstivity
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to measure small leaks, will identify the leakage source to the extent practical, and will be provided with suitable control room alarms and readouts. The leakage detection systems proposed to detect leakage from components and piping of the reactor coolant pressure boundary follow the recommendations of NRC Regulatory Guide 1.45 as far as practical and 3
thus provide reasonable assurance that any structural degradation f
resulting in leakage during service will be detected in time to permit
. corrective actions. This constitutes an acceptable basis for satisfying the requirements of'NRC Ceneral Design Criterion 30, Appendix A of 10 CFF Part 50.
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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 considerations that make it necessary to consider potential vessel failure for WPPSS-Nuclear Projects 3 and 5.
The bases for our conclusion are that the design, material, fabrication, inspection, and quality assurance requirements for the reactor vessels of WPPSS-Nucicar Projects 3 and 5 will conform to the rules of the ASME Code, Section 111,.1971 Edition, and the 1972 Summer Addenda. Alsu, escrating.
limitations on temperature and pressure will be established for this plant in accordance with Appendix G, " Protection Against Non-Ductile Failure," of the 1972 Summer 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:
a.
Will be designed and fabricated to the high standards of quality required by the ASME Code Section III and pertinent Code Cases listed above.
b.
Will be made from materials of controlled and demonstrated high quality.
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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 during most upsets in operation, and that the vessel will not fail under the conditions of any of the postulated accidents, e.
Will be subjected to monitoring and periodic inspection to demonstrate that the high initia] quality of the reactor vessel has not deteriorated significantly under the service conditions.
f.
Itay be annealed to restore the material toughness properties if this becomes necessary.
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COMPONENT AND SUBSYSTDI DESIGN Steam Cencrator Tube Integrity We have evaluated the factors that affect the integrity of the steam generator tubes for WPPSS Units 3 and 5.
We conclude that reasonable measures have been taken to ensure that the tubing will not be subjected to conditions that will cause deleterious vastage or cracking. Our conclusion is based on the following:
1.
The steam generators will be of advanced design with improved secondary water flow characteristics. This will provide more tolerance for occasional lack of control of the secondary water chemistry.
2.
All volatile treatment is planced for secondary water chemistry control, thereby minimizing the probability of deleterious local
.high concentrations of caustic or phosphate on the tubing.
3.
The condenser will be made with Ni-Cr-Fe alloy tubing, thus minimizing the probability of condenser leakage contributing to contamination of the secondary water.
4.
We will require inservice inspections that reficct the criteria of j
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Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water t
3 Reactor Stean Cencrator Tubes."
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i COMBUSTION E;GINEERING, L';CORPORATED COMBUSTION E;GINEERING STA';DARD SAFETY A';ALYSIS REPORT (CESSAR)
DOC 1'ET NO. 50-470 SAFETY EVALUATION 11ATERIALS ENGINEERING BRANCil, L PERFORMANCE SECTION REACTOR COOLA!.T SYSTDI A';D CONNECTED SYSTE!S Integrity of Reactor Coolant Pressure Boundary
- Fracture Touchness 1.-
Compliance trith Code Reouirements We have reviewed the caterials selection, toughness requircaents, and extent of r.aterials testing proposed by the applicant to provide assurance that the ferritic naterials used for pressure retaining components of the reactor coolant boundary will have adequate tough-ness under test,. normal operation, and transient conditions. The ferritic natorials arc specified to nect the toughness requirements of the ASME Code,Section III, inclu. ding Sureer 1972 Addenda. In addition, natcrials for the reactor vessel are specified to nect the additional test requirer.cnts and acceptancd criteria of Appendix G, 10 CFR 50.
The fracture toughness tests and procedures required by Section III of the ASME Code, as aug=cnte'd by Appendiv. G, 10 CFR 50, for the reactor vessel, provide reasonable assurances that adequate safety margins against the ' ossibility of nonductilo behavior or rapidly p
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1 propagating' fracture can be established for all pressure retaining components of the reactor coolant boundary.
2.
Operating Linitations The reactor will be operated in accordance with the ASME Code,Section III, including Su=mer 1972 Addenda, and Appendix G,10 CFR 50.
This will minimize the possibility of failure due to a rapidly propagating crack. Additional conservatism in the pressure-temperature limits used for heatup, cooldoun, 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 AEC Regulations, will ensure adequate safety margins during operating, testing, maintenance, and postulated accident conditions. Compliance with those Code provisions 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 Staterial Surveillance Progran The toughness properties of the reactor vessel beltline material will be monitored throughout service life with a material surveillance 4
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. program that will meet the requirements of ASTM Specification T
E 185-73. This program also complies with Appendix II,10 CFR Part 50 c
except that specimen holders are attached to the vessel cladding.
Combustion Engineering has submitted a Topical Report CENPD-155P, "CE Procedure for. Design, Fabrication, Installation and Inspection of Surveillance Specimen Holder Assemblics." We have evaluated this
, report and conclude that their method of attaching capsule holders to
.the' vessel clad is acceptabic and results in no degradation of the vessel base material.
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 above doc-uments are n.ct. ' Compliance with these documents will ensure that the
'i surveillance program consistutes an acceptabic basis for monitoring radiation induced changes in the fracture toughness of the reactor vessel material, and will satisfy the requirements of AEC Cencral DesignCriterion31,AppendixAof10'CFIpPart50.
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 prorcrties, the applicant has stated that should results of tests indicate that the toughness is not adequate, the reactor vescel can be annealed to restore the toughncos to acceptabic levels.
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t Reactor Vessel Closure Studs _
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The reactor vessel closure studs will be procured and initially inspected in conformance with the mechanical and toughness property requirements and inspection requirements recommended in Regulator Studs."
Getide 1.65, " Materials and Inspections for Reactor Vessel Closure However the inservice inspection requirements are not satisfactory.
Inservice inspection requirements of reactor vessel studs will be b
e basis.
covered in Applicant's SAR and will be reviewed on a case-y-cas 5.
Inservice Inspection Program l td To ensure that no deleterious defec'ts develop during service, se ec e i dically..The welds and weld heat-affected zones will be inspected per o tem applicant has stated that the design of the reactor coolant sys d
incorporates provisions for access for inservice inspections in accor -
Vessel Code, and
-ance with Section XI of the ASME Boiler and Pressure h remote l'
that suitabic equipment will be developed to facilitate t e ible inspection of those areas of the reactor vessel not readily access The conduct of periodic inspections and hydro-to inspection personnel.
oolant static testing of pressure retaining compo,nents in the reactor c ASME Section pressure boundary in accordance with the requirements of 1
XI Code provides reasonable assurance that evidence of structur,a degradation or loss of Icaktight-integrity occurring during service f ty will be detected in time to palmit corrective action before the sa e Compliance with the inservice function of a component is compromised.
for innpcetions required by this Code constitutes an acceptable basis l
32, satisfying the requirements of AEC Cencral Design Criterion Appendix A of 10 CFR Part 50.
L.
To ensure that no deleterious defects develop during sc vice in ASME Code Class 2 system components, selected welds and wcld heat-affected zones are inspected prior to reactor startup and periodically throughout the life of the plant. Code Class 2 systens 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.
Exampics of Code Class 2 systems are:
(1) Residual heat removal systems, (2) Portions of chemical and volume control systems, and (3) Engineered safety features not part of Code Class 1 systems.
Exampics of Code Class 3 systems are:
(1) Component cooling water systems and (2) Portions of raduaste systems. All of these systems transport fluids. The applicant has stated that the Code Class 2 systems meet the requirements of ASME Section XI.
The Code Class 2 systems and Code Class 3 systems arc in conformance with the recom-mendations of Regulatory Guide 1.51.
Comi;11cncewiththeinservice inspections required by this Code and Regulatory Guide constitutes an acceptablc bases for satisfying AEC Concral Design Criteria 36, 39, 42, and 45, Appendix A of 10 CFR Part 50.
6.
Pump Plyyheci The probability of a loss of pump flywhcci integrity can be minimized by the use of suitable material, adequate design,'and inservice inspection.
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The applicant has stated'that the integrity of the reactor coolant pump flywhccl'is provided by compliance witfi the AEC Regulatory Guide 1.14 " Reactor Coolant Pump Flywhccl Integrity."
The use'of suitable material, and adequate design and inservice inspection for the flywhecis of reactor coolant pump motors as specified in ~the SAR provides reasonable assurance (a) that the structural integrity of the flywhccls is adequate to withstand the forces imposed in the event of pump design overspeed transient without loss of function, and (b) that their integrity will be verified pcriodically in service to assure that the required level of soundness of the flywhcci materini is adequate to preclude failure.
Compliance with the recommendations of AEC Regulatory Guide 1.14 constitutes an acceptable basis for satisfying the requirements of AEC Cencral Design Criterion 4, Appendix A of 10 CFR Part 50.
7.
RCPB Leakage Detection System Coolanticahagewithintheprimarycontaint$cntmaybeanindication of a smalt. through-vall fl'w in the reactor coolant pressure boundary.
a The leakage detection system. proposed for leakage to the containment will include diverse leak detection methods, will have sufficient sensitivity to measure cmall leaks, will identify the leakage source to the extent practical, and will be provided with s~uitabic control room alarma and readoute.. The Icakage detection systems proposed to i
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1 detect leakage from components and piping of the reactor coolant pressure boundary are in accordance with AEC Regulatory Guide 1.45 insofar as it is applicabic to CESSAR and thus provide reasonabic assurance that any structural degradation resulting in 1cakage during service will be detected in time to permit corrective actions.
Compliance with the recommendations of AEC Regulatory Guide 1.45 constitutes an acceptabic basis for satisfying the requirements of AEC Cencral Design Criterion 30, Appendix A of 10 CFR Part 50.
8.
Reactor Vessel and Appurtenances Reactor Vessel Integrity:
L'e have revicued all factors contributing to the structural integrity of the reactor vessel, and we concludu there are no special considera-tions that make it necessary to consider potential vessel failure for CESSAR.
The bases for our conclusion are that the design, material, fabrica-tion, inspection, and quclity assurance requirements for the reactor vessels of Units 1 and 2 will conform to ti c rules of. the ASME Code,Section III, 1971 Edition, and the 1972 Summer Addenda. Also, operating limitations on temperature and pressure will be established for this plant in accordance with Appendix G, " Protection Against Non-Ductile rallure," of the 1972 Summer Addenda of the ASMC Boiler and Prcscure Vessel Code,Section III, and Appendix G, 10 CFR 50.
)
The integrity of the reactor vessel is assured because the vessel:
Will be designed and fabricated to the high standards of quality a.
required by the AS:tE Code Section III and pertinent Code Cases listed above.
b.
Will be cade from materials of controlled and demonstrated high quality.
Will be inspected and tested to provide substantial assurance c.
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 vecsel design conditions will not be exceeded during normal reactor operation or during most upsets in operation, and that the vessel vill not fail under the conditions of any of the postulated accidents.
Uill be subjected to monitoring and periodic inspection to c.
demonstrate that the high initial quality of the reactor vessel has not deteriorated significantly under the service conditions.
f.
!!ay be annealed to restore the material toughness properties if this becones necessary.
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l R\\TERIALS PERFOL'Li';CE SECTION, MATERI Al.S ESCINEERING BRANCH REFERE::CES Ceneral 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, "AEC codes and Star.6ard Rules -
Applicabic Codes, Addenda, and Code Cases "In Effect" for Conponents that are part of the Reactor Coolant Pressure Boundary," June 12, 1971.
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" Standard Forcat and Content of Safety Analysis Reports for Nuc1 car Perer Plants," Rev. 1, October 1972.
Fracture Tough.iens 10 CFR 50 - Appendix G, " Fracture Toughacss Requirctents," June 1,1973.
ASNE Boiler and Pressure Vessel Code,Section III,1972 Su:=cr Addenda, including Appendix G, " Protection Against Non-Ductile Failure."
AS:!E Specification, SA-370-71b, " Methods and Definitions for Mcchanical Testing of Stect Products," ASME Boiler and Pressure Vessel Code,Section II, Part A - Ferrous, 1971 Edition, Summer and Winter, 1972 Addenda.
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- 2-AbT.! Specific., tion E-208-C9, " standard 3fethod for Conducting Dropwoight Test t Deternine Nil-Duetility Transitica Temperature of Ferritic Sceals," Annual Eoc'( of ASTM Standards, Part 31, July 1973.
ASTU Specification E 23-72, "hotched Bar impact Testibg of Met.illic Materlah," Annual Book of.'.STM Standards, Part 31, July 1973.
Material Survaillarcc Pror,rans 10 CIR 50 - Appendix H. " Reactor Vessti Material Surveillance Prcgram Requircecnts," Juac 1, 1973.
t ASri specification E-itS-73, " Surveillance Testr or Structural ;faterials in Mu;1 car Ecacters," Annual Book of ASTM Standards, Part 30 Joly 1973 fupp Plywheels AEC Regulatory Guide 1.14 "reacter Coolant Pucp Flywhccl lutegrity,"
October 27, 1971.
RCPD LeaLace Detection Systems
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AEC Regulatory cuide 1.45, " Reactor Coolant Pre'ssure Boundary Leakage Detection Systens," May 1973.
Inservice Innpcetinn Program AEC Cuideline Document, " Inservice Inspection Requirements for Nuc1 car Power Plantr. Constructed with Limited Accessibility for Inservice Inspections," January 31, 1969.
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3-ASME Loller and Pccssure Vessel Code Section XI, 1971 Edition, 2ncluding Winter 1971, Sunner 1971, Winter 1972, and Summer 1973 Addenda.
Regulatory Guide 1.51, " Inservice inrpection of ASMF, Class 2 and 3 Nuc1 car Poter Plant Com,ionents," !!ay 1973.
Reactor Ver.sel Intetrit,v ASME holler and Pressure Vcssel Code,Section III,1971 Edition pli:a Addenda thtCJ3h Winter 1972.
ASME Boiler and Pressore Vessel Cede,Section XI, 1971 Edition plus AddenJa cl. rough Winter 1972.
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