ML19220B940
| ML19220B940 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/14/1975 |
| From: | Maccary R Office of Nuclear Reactor Regulation |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904270681 | |
| Download: ML19220B940 (12) | |
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50-320 MS 24-12 Voss A. Moore, Assistant Director for Light Water Reactors, Crcup 2 Division of Re.'etor Licensing ETROPOLITAN EDISCN CCMPANY, TH2EE HII2.SLAND, UNIT 2 (CL), DOCKE" huGZR 50-320 Plant.Name: Three Mi.le Island, Unic 2 Suppl.iers : Babcock & Wilcox; Burns and Roe Licensing Stage: CL Docket Number: 50-320 Responsible Branch 'ed Project Manager: LWR 2-2; H. Silver Requested Compir ' Date: October 10, 1915 Task: SER Review Statt Coc:plete The information submitted by the applicant, including Amend =ent No. 32 has been reviewed by the Mate +1sla Branch Office of % clear Reactor Ecgulation.Perfornance Section, Materials Engineering Our sections of the Safety Evelustion are snelnsed.
R. R. Maccary, Assistant Director for Engineering Division of Technfcal Reviev Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enc 1:
D. Pfeank"c, NRR cc w/o enc 1:
W. C. l'e. Donald, MI?C R. E. Fein -an, n R. Boyd, HL S. Varga, KL H. rabe -'yi, TR I. Iniel, RL H. Silver, EL Distribution:
S. S. Pav11cH, TR g r-File 50-320 W. S. Hazelton, TR U. Petapovs, TR NRR k g.
H. 7. Conrad, TR MTE3 Rdg.
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qqp METROPOLITMi EDISON CCMPA'"?
THREE MILE ISLAND NUCLEAR STA!!CN C'i!T 2 (CL)
DCCRET :D13ER 50-320 SAFETY EVALUATICN MATERIALS ENGINEERING 3 RANCH MATERIAI.S PERFORMANCE SECTICN REACTOR COOLMIT SYSTDi AND CCNNECTED SYSTDiS Integrity of Reactor Ccolant Pressure Boundarv Fractura Touchness 1.
Compliance with Code Recuire=ents We have reviewed the materials select.on, toughness require =ents, and extent of =aterials testing accceplished by the applicant to provide assurance that the =aterials used for the reactor vescel vill have adequate toughness under test, normal operation, and transient conditions. All =aterials =et the toughness requirements of the ASMI Boiler and Pressure Vessel Code,Section III, 1965 Edition and Addenda through Su=ner 1967.
All the tests required by Appendix G, 10 CFR Part 50 have not yet been conducted, however, we vill require that the ferritic =aterials used for the reactor pressure vessel meet the requirements of Appendix G, 10 CFR 50, as far as practical. The fracture toughness tests and procadures required by Section III of the ASME Code, and Appendix G, 10 CIR 50, for the reactor vessel, provide reasenable assurance that adequate safety =argins against the possibil,ity of nonductile behavior or rapidly propagating fracture can be established for the pressure-retaining components of the reactor coolant boundary.
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Ocerating L1=itations '
The reactor will be operated in a =anner that will minimize the
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possibility of rapidly propagating failure, in accordance with Appendix G to Section III of the ASMI Boiler and Pressure Vessel Code and Appendix G, 10 CFR 50.
Additional conservatism in the pressure-temperature limits used for heatup, cooldcwn, testing, and core operation will be provided because these will be deter =ined 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 li=itations, using results of the fracture toughness tests perfor=ed in accordance with the Code and NRC Regulations, will ensure adequate safety =argins during operating, testing, maintenance, and postulated accident conditions. Co=pliance with these Code provisions and NRC regulations constitute an acceptable basis for satisfying the requirements of NRC General Design Criterion 31, Appendix A of 10 0FR Part 50, 3.
Reactor Vessel Materials Surveillance Procram The toughness properties of the reactor vessel beltline =aterial will be =onitored throughout service life with a caterial surveillance program. This progras is described in Topical Report 3r4-10100,
" Reactor Vessel Material Surveillance Program," which we have reviewed and found acceptable. The program =eets the require =ents of ASIM E 185-73 and Appendix H,10 CFR 50 (July 17,1973), as far as practical.
Changes in the fracture toughness of =aterial in the reactor vessel beltline caused by exposure to neutron radiation will be assessed 84-100
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properly as we will require that the RT NDT caterial will be esti=ated according to the guide lines of Regulatory Guide 1.99 instead of the topical report 3AW-10056A. Adequate safety
=argins against the possibility of vessel failure yill be provided since the essential caterial surveillance require =ents of ASTM E 185-73 and Appendix 'd, 10 CFR Part 50, are met.
The surveillance progra=
constitutes an acceptable basis for monitoring radiation induced changes in the fracture toughness of the reactor vessel =aterial, and will satisfy the require =ents of NRC Ceneral Design Criterion 31, Appendix A, of 10 CFR Part 50.
4.
Pu=n Flywheel The probability of a loss of pu=p flywheel integrity can be =ini=ized by the use of suitable =aterial, adequate design, and inservice inspection.
The integrity of the reactor coolant pu=p flywheel is provided by having it designed for a 114% overspeed condition. In ad'ition, a 1000 ultrasonic volumetric inspection of the flywheel plate before machining using ASME Section '!I acceptance criteria, was performed.
Inservice inspections of the flywheel vill be performed in accordance with the reco==endations of NRC Regulatory Guide 1.14, as far as practical.
We conclude that the provisions for =aterial selection and flywheel design, and inservice inspections in accordanca with Regulatory Guide 1.14 ensure adequate flywheel integrity and constitutes an acceptable basis for satisfying the requirements of NRC General Design Criterion 4, Appendix A, 10 CFR Part 50.
8C101
_4 REACTOR VESSEL AND APPURTENACES Reactor Vessel Inte2rity We have reviewed all factors contributing to the structural integrity of the reactor vessel and we conclude that the vessel is not likely to fail as we will i= pose sufficient restrictions on the operational li=its of pressure and temperature after requiring that 1.
The upper shelf C value for the noz:le belt =aterial will be v
deter =ined.
2.
The RT f the reactor vessel welds be suitably esti=ated, the NDT charpy value for these welds has been estimated to be 66 f t.
lbs.
before irradiation and the Cu content of the middle circu=ferential weld is.29%
Otherwise, the design, material, fabrication, inspection, and quality assurance requirements conformed to the rules of the ASME Boiler and Pressure Vessel Code.Section III, 1965 Edition, all Addenda through Sc==er 1967, and all applicable Code Cases.
Operating li=itations on te=perature and pressure are established for this plant in accordance with Appendix G, " Protection Against Nonductile Failure," of the 1972 Su==er 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.
Was 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.
Was =ade fro naterials of controlled and demonstrated high quality.84-102
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Was extensive 17 inspected and tested to provide substantial assurance that the vessel will not fail because of =aterial or fabrication deficiencies.
4.
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.
5.
Will be subjected to monitoring and periodic inspection to de=onstrate that the high initial quality of the reactor vessel has not deteriorated significantly under the service conditions.
Inservice Inspection Program To ensure that no deleterious defects develop during service, selected welds and weld heat-affected zones will be inspected periodically. The applicant has stated that the reactor vessel, steam generators, and pressurizer have been designed and arranged, to the extent possible to permit compliance with subsections IS-141 and 15-142 of Section XI of the ASME Code, " Inservice Inspection of Nuclear Reactor Coolant Systems."
Access has been provided to perfor: volumetric examinations of the pri=ary side pressure-containing welds fro = the external surfaces of these vessels.
We will recuire that they follow all the crovisions of ASME Ccde,Section XI.
The conduct of periodic inspections and hydrostatic testing of pressure-retaining co=ponents in the reactor coolant pressure boundary in accordance 81^103
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with the require =ents of ASME Section XI Code provides reasonabl.
assurance that evidence of structural degradation or loss of leaktight-integrity occurring during service will be detected in ti=e to per=it corrective action before the safety function of a component is co=prc=1 sed.
Co=pliance with the inservice inspections required by this Code constitutes an acceptable basis for satisfying the requirements of :!RC General Design Criterion 32, Appendix A of 10 CFR Part 30.
RCPB Leakage Detection Systq=
Coolant leakage within the contain=ent =ay be an indication of a s=all through-wall flaw in the' reactor coolant pressure boundary.
The leakage detection syste=s proposed for leakage to the containment includes diverse leak detection =ethods, has sufficient sensitivity to
=easure s=all leaks and can identify the leakage source to the extent practical. The major co=ponents of the syste= are the containment vessel su=p and radiogas and air particulate radioactivity =onitors.
Intersyste=
leakage will be detected by abnor=al readings frc= radioactivity =onitors in the secondary syste=.
The construction per=it for this plant was issued before the issuance of Regulatory Guide 1.45, however, the applicant =eets the important requirements of the Regulatory Guide 1.45.
The =ain exceptions are that syste=s are not specifically seis=ically qualified and all syste=s do not have a con-trol roo= readout. Both of these requirements were established subsequent to_the SAR application date. The leakage detection syste=s proposed to detect leakage frc components and piping of the reactor coolant pressure 84'104 boundary are generally in accordance with NRC Regulatory Guide 1.45 and provide reasonable assurance that any structural degradation re-sulting in leakage during service will be detected in ti=e to permit corrective actions. This da3_'e of conformance with the recccmendations of NRC Regulatory Guide 1.45 constitutes an acceptable basis for satisfying the requirements of NRC General Design Criterion 30, Appendix A of 10 CFR Part 50.
CCMPONE:;T AND SUBSYSTEM DESIC'T Steam Generator Tube Integrity We have evaluated the factors that affect the integrity of the steam generator tubes for Davis-Besse 1.
We conclude that all reasonable measures have been taken to ensure that the tubing vill not be subjected to conditions that will cause deleterious wastage or c acking. Our conclusion is based on the following:
1.
There have been no instances of tube degradation in steam generators of the once through design that will be used in the Three Mile Island Unit 2 plant.
2.
The secondary water chemistry control used will be all volatile, thereby
~4-4 4-ing the probability of deleterious local high concentrations of caustic or phosphate on the tubing.
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3.
To further control i= purities in the secondary water to very 1cw levels, Three ile Island Unit 2 will use full flow de=ineralization of the condensate.
4 We will require that periodic inservice inspections of the steam
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generator tubes be performed in accordance with the reco==endations of the revision to Regulatory Guide 1.83, " Inservice Inspection of Pressuri:ed 'a'ater Reactor Steam Generator Tubes."
TECHNICAL SPECIFICATIONS The Technical Specifications have not been submitted for review. They will be reviewed when available.
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PATERIALS E' GI::EERI::C 3?X:CII MATERIALS PERFCF20.':CE SECTIO::
REFERENCES General Federal Recister 10 CFR Part 50, Appendix A, " General Design Criteria Nuclear Plants," July 7, 1971.
Federal Recister 10 CFR Part 50 5 50.55i "u: Codes and Standard Rules -
Applicable Codes, Addenda, and Code Cases 'In Effect' for Cc=ponents that are Part of the Reactor Coolant Pressure Eaundary," February 15, 1974
" Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 1, October 1972.
General Materials ~Censiderations Materials Reecificatiens ASMI Betler and Pressure Vessel Code,Section III, 1965 Edition, plus
!.ddenda through Sc==er 1967.
a.
Paragraph :i3-2121: Permitted Material Specificatiens b.
Paragraph N3-2122: Special Requirements Conflicting with Permitted Material Specifications.
List of NRC Accroved Cede Cases, February 22, 1973.
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Fracture Touchness 10 CFR 50 - Appendix G, " Fracture Toughness Requirements," July 17, 1973.
ASME Boiler and Pressure Vessel Code,Section III, 1972 Su==er Addenda, including Appendix 0, " Protection Against Ncn-Ductile Failure."
ASTM Specification E 208-69, " Standard Method for Conducting Dropweight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels," Annual Book of ASTM Standards, Part 31, July 1973.
Material Surveillance Programs 10 CFR 50 - Appendix H, " Reactor Vessel Material Surveillance Progras Requirements," July 17, 1973.
ASTM Specification E 185-73, " Surveillance Tests on Structural Materials in Nuclear Reactors," Annual Book of ASTM Standards, Part 30, July 1973.
Pu=o Flvwheels NRC Regulatory Guide 1.14, " Reactor Coolant Pump Flywheel Integrity,"
October 27, 1971.
RCPB Leakage Detection Systems.
NRC Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.
e 84 108
Inservice Inscection Procram 1.
NRC Guideline Docu=ent, " Inservice Inspection Requirements for Nuclear Power Plants Censtructed with Limited Accessibility for Inservice Inspections," January 31, 1969.
2.
ASME Boiler and Pressure Vessel Code,Section XI, 1974 Edition.
3.
Regulaccry Guide 1.51, " Inservice Inspection of ASME Class 2 and 3 Nuclear Power Plant Components," May 1973.
Reactor Vessel Inteerity 1.
ASMI Boiler and Pressure. Vessel Code,Section I!!, 1965 Edition plus Addenda through Su==er 1967.
2.
ASMZ Boiler and Pressure Vessel Code,Section XI, 1974 Edition.
Steam Generator Tube Integrity NRC Regulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steca Generator rubes," June 1974.84-109 Me
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