ML19350A474
| ML19350A474 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/06/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Gary R TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| References | |
| NUDOCS 8103160261 | |
| Download: ML19350A474 (25) | |
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Docket tios. 50-445 MAS 0IS01 and 50-446
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Mr. R. J. Gary
'o Executive Vice President and i
General Manager 4f4g l [1 B
Texas Utilities Generating Company j~
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O g8 2001 Bryan Tower y'4hg,bf
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Dallas, Texas 75201
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Dear Mr. Gary:
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SUBJECT:
REQUEST FOR ADDITI0f1AL If1FORMATI0fl FOR COMAfiCHE PEAX STEA STATIO:4, Uti!TS 1 A?iD 2 The Component Integrity Section, Materials Engineering Branch, Division of Engineering has reviewed the Final Safety Analysis Report (FSAR) for Comanche Peak Steatr. Electric Station Units 1 and 2.
Based on our review of this information, we have prepared a draft input to the Safety Evaluation Recort.
In this draft to the Safety Evaluation Report we have identified areas for which sufficient information has not been submitted to determine compliance with or justify exemptions to Appendices G and H,10 CFR Part 50. The FSAR also must include additional infonnation to define reactor vessel temoerature-pressure limits, ensure pump flywheel integrity and ensure fracture prevention of the containment pressure boundary. The areas where sufficient information has not been provided will remain open items until Texas Utilities Generating Company provides the necessary information.
We are enclosing the draf t to the Safety Evaluation Report (Enclosure 1) to assist you in understanding our requirements for additional infomation and to ensure a complete and adequate response. We have listed our specific recuire-ments in a Request for Mdi;1onal Information (Enclosure 2). Please amend your FSAR to include the information requested in Enclosure 2 Please expedite your response to this request for additional information in order that we may resolve these open items prior to the issuance of the Safety Evalua-tion Report.
We also request that you keep us advised on the schedule for submittal of this information to assist in our management of staff resources.
8108160 N
Mr. R. J. Gary.YAR 61331 Should you have questions concerning tnis request #or acL'tional informatitn or desire ta meet with us on tnis matter, please contact us.
Sincerely, h
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Rocert L. Tedesco Assistant Director for Licensing Division of Licensing
Enclosures:
1.
Draft Input to the Safety Evaluation Report 2.
Request for Additional Information cc w/ enclosures:
See next page 4
Mr. R. J. Gary Executive Vice President and General FMnager MAR 61331 Texas Utilities Generating Company 2001 Br Dallas,yan Tower Texas 75201 4
Nir.holas S. Reynolds, Esq.
Mr. Richard L. Fouke cc:
Detevoise & Liberman 1200 Seventeenth Street Citizens for Fair Utility Regulation 16CS-B Carter Drive Washington, D. C.
20036 Arli5gton, Texas 76010 J.
Spenc~er C. Relyea, Esq.
Worsham, Forsythe & Sampels Resident Inspecter/Cc=anche Peak 2001 Bryan Tower Nuclear Power Station
'e Dallas, Texas 75201 c/o U. S. Nuclear Regulatory Ccanission P. O. Box 33 Mr. Homer C. Schmidt Glen Rose, Texas 76043 Manager - Muclear Services Texas Utilities Services, Inc.
2001 Bryan Tower Dallas, Texas 75201 Mr. H. R. Rock Gibbs and Hill, Inc.
393 Seventh Avenue New York, New York 10001 Mr. A. T. Parker Westinghouse Electric Corporation P. O. Box 355
. p, Pittsburr4, Pennsylvania 15230
. David J. Preister Assistar.t Attorney General Env.onmental Protection Division
,1 P. 0. Box 12548, Capitol Station
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Mrs. Juanita Ellis, President Citizens Association for Sound Energy
,1426 South Folk
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Dallas, Texas 75224 Geoffrey M. Gay, Esq.
West Texas Legal Services 100 Main Street (Lawyers Bldg.)".
Fort.W.o.r.t.h., T_exas 76102 7 '
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t ENCLOSURE i CCMANCHE PEAK STEAM ELECTRIC STATICN UNIT NUM5ERS 1 AND 2 CRAFT INPUT TO THE SAFETY EVALUATION REPORT MATERIALS ENGINEERING BRA k H COMPCNENT INTEGRITY SECTION 5p3.1 Reactor Vessel Materials General Design Criterion 31, " Fracture Prevention of Reactor Ccolant Pressure or -
Boundary," Appendix A,10 CFR Part 50, requires, in part, that ne reactor coolant pressure boundary be designed with sufficient margin to ensure that, when stressed under operating, maintenance, testing, and postulated accident conditions, the boundary behaves in a nonbrittle canner and the probability of rapidly propagating fracture is minimized. General Design Criterion 32,
" Inspection of Reactor Coolant Pressure Ecundary," Appendix A,10 CFR Part 50.
requires, in part, that the reactor coolant pressure boundary be designed to permit an appropriate material surveillance program for the reactor pressure boundary. Materials selection, toughness requirements and extent of material testing were reviewed in accordance with the above criteria subject to the r_les and requirements of 10 CFR Part 50 Paragraph 50.55a " Codes and Standards",
10 CFR Part 50 Appendix G " Fracture Toughness Requirements", and 10 CFR Part 50 Appendix H " Reactor Vessel Materials Surveillance Program Requirements.
The Comanche Peak Units 1 and 2 construction permits were received in Dece:ter 1974. Based upon tne construction permit date,10 CFR Part 50 Paragra;n 50.55(a) requires that ferritic materials used for vessel in tne reactor coolant pressure boundary be constructed to Section III of the ASME Code no earlier than tne Summer -72 Addenda of the 1971 edition and that ferritic materials used for pressure retaining piping, pu=p and valve ccroonents in the reactor ccolant pressure boundary be constructed'to Section III of the ASME Code no earlier
. than the Winter 72 Addenda of the 1971 edition. Ferritic materials used for fabrication of pressure vessels, piping, pump and valve components that are part of the reactor coolant pressure boundary were constructed to ASME Code Addenda which satisfy the requirements of 10 CPR Part 50 Paragraph 50.55(a).
/
Comoliance with Accendix G,10 CFR Part 50 We have evaluated the applicant's FSAR to detemine the degree of compliance e-with fracture toughness requirements of Appendix G,10 CFR Part 50. Our evaluation indicates that the applicant complied with Appendix G,10 CFR Part 50, except for Paragraphs I.A. I.C, III.B.4, IV.A.1, IV. A.3, I'i.A.4 and IV.3, which will remain open items until the applicant submits the requested data. Our evaluation of each of these areas follows.
Paragraph I.A states that the adequacy of the fracture toughness of ferritic materials that are used in the reactor coolant pressure boundary and have a minimum specified yield strength greater than 50,000 psi te demonstrated to the Comission on an individual case ba1 is. Table 5.2-2 of the Comanche Peak Steam Electric Station FSAR indicates that SA-533 Grade A Class 2 steel has been used in the reactor coolant pressure boundary.
SA-533 Grade A Class 2 steel has a specified minimum yield strength of 70,000 psi. The applicant mus'. supply fracture toughness data from a sufficient number of heats to demnstrate the generic fracture tougnness of SA-533 Grade A Class 2 steel.
Westinghouse' Topical Report WCAP-9292, "Dyrmic Fracture Tougnness of ASME SA-508 Class 2 and ASME SA-533 Grade A Class 2 Sase and Heat Affected Zone Material and Applicable Weld Metals", contains fracture toughness da:a frcm a sufficient numoer of heats of,SA-533 Grade A Class 2 steel to demonstrate tne generic fracture toughness of SA-533 Grade A Class 2 steel. The Westincnouse m
. report indicates that the conclusions concerning use of hign strengtn materiais in the reactor c.colant pressure boundary are appifcable to Comanene Peak Units 1 & 2.
The apolicant must revise the~ FSAR--to indicate that :ne conclusions in
)
Westinghcuse Topicil Report '4 CAP-9292 are applicable to Cemanche Peak Units 1 and 2 SA-533 Grade A Class 2 steel.
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Paragraph I.C requires the applicant to demonstrate to the Ccmission on an individual case basis the adequacy of fracture toughness in any ferritic bciting or fastener hav'.ng a specified yield strength greater than 130 ksi.
The applicant has indicated that closure bolting of SA-540, 3A-320, SA-ig3 or SA-453 Grade 660 material was used in the reactor ccolant pump.
SA-540 Class 1 and 2 material specifies a minimim yield strength greater than 130 ksi.
If SA-540 Class 1 or 2 material was used in the reacter coolant pumo the applicant must demonstrate the adequacy of its fracture toughness.
Paragraph III.S.4 requires' individuals cerformino fracture touchness tests be qualified by training and experience and that individual demonstrate ccmpetency to perfom tests in accordance with a written procedures.
The appli: ant has not provided any information that demonstrates compliance with these fracture toughness tests requirements.
The applicant must provide the required information or submit another method a
of qualifying personnel wnich is equivalent to the requirements of Paragrap III.3.4.
?aragraph IV.A.1 requires that all ferritic material used in vessels that are part of the reactor ecolant pressure boundary be tested to the requirerents of I
4 Paragraph NB-2300 of the ASME Code. Paragrapn NS-2330 of :ne Summer 1972 Addenda to the ASME Code requires a reference tem:erature, RT
.to be determined NDT for all base metal, weld metal and heat affected :ene material. The reference temperature is determined using the results of botn the drop weign: test and the Charpy V-notch test. Longitudinal and transverse enarpy V-noten tests y
are required for all beltline vessel base metal by Paragraph NS-2330. Trans-verse charpy V-r.otch tests are required for all reactor vessel materials out-ar-side tne beltline region, for weld metal and heat affected 7nnas in the beltline region, for primary side steam generator materials, and for pressurizer ma terials. Drop weight tests are required for all weld metal and base metal in the reactor vessel, pressurizer and the primary side of the steam generator. The results of the drop weight tests on the adjacent base metal can be used in establishing the reference temperature of the heat affected zone. The applicant has reported the reference temperature and drop weight test results for reactor vessel base metal, but has not supplied longitudinal and transverse charpy V-notch test results.
The applicant has reported the reforence temperature and drop weignt test results for belt'.ine welds. tut has not supplied transverse charpy V-notch test results.
The applicant has not reported the reference temperature and the transverse charpy V-notch test results for any heat affected :ene in the reactor vessel, the pressuri:er, or the primary side e,f the steam generator. The applicant has not reported the reference temperature, transverse impact properties, and drop weight test results for weld metal in the pressurizer, the primary side of the steam generato: and the reactor vessel outside the beltline region.
5 The applicant has not reper:ed the reference ter:erature, transverse ir:ac:
properties, and drop weign: tes results for base retal in :ne ;res:uri:er and the primary side of :ne steam generater.
Before our review of the applicants compliance with Paragra:n IV.A.1 can be completed, the applican =ust supply the above icentified missing inf:rmatien.
S Pare,raph IV.A.3 requires ferritic materials for piping, purcs, anc va?ves
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that are part of the reacter coolant pressure bcuncary to meet tne fracture m$'
toughness requirements of Paragraph N5-2332 of thi',5ME Code. Table 5.2-2 of FSAR indicates ferritic caterials were used for reactor coolant piping and either ferritic cr austenitic material was used f:r reac:cr coolan; pressure boundary valve disks. The applicant has net proviced any fracture ::ugnness data fer ferritic materials used for reac:ce c: alan: piping and reac r ecolant pressure boundary valve disks. The applicant must submit fracture toughness data for all ferritic materials used for reac r c:olant pressure boundary piping and reactor coolant pressure boundary valve disks to terc.' strate ccepliance with Paragraph IV.A.3.
Paragraph IV.A.4 has been revised to require ferritic fasteners wnich are used in reactor c olant pressure boundary application and that nave a nc=inal diameter exceeding (1) inch to be Charpy V-notch impact tested ace:rding to the require-ments of NS-2333 of the ASME Code. Table 5.2-2 of the FIAR incicates ferritic materials were used for reactor coolant pressure boundary fasteners in the reactor vessel, the steam generator, the pressuri:er, the reac::r coolan: pcr:s a
and the reactor coolant pressure boundary valves. The applicant has su::ittec char y V-notch impact data for ferritic fasteners used in :ne reac:cr vessel.
. No charpy V-r.otch impact data has been submittee for any other ferritic fastener used for reactor coolant pressure boundary applications. To demonstrate that reactor coolant pressure boundary ferritic fasteners comply with minimum fracture toughness requirements of Paragraph IV.A.4 provide charpy V-notch impact test data for all reactor coolant pressure boundary ferritic fasteners.
j Paragraph IV.B requires unirradiated reactor vessel beltli.'* "aterials frcm base metal, weld =etal and heat affected zone samples to have a minimum upper shelf charpy V-notch er.ergy of 75 ft-lbs.
The applicant has reported that the upper shelf energy for all unirradiated beltline weld metal and base metal exceeds 75 ft-lbs. The applicant has not submitted the actual charpy V-notch impact test results.
The applicant has not submitted any infor ation concerning the upper-shelf energy level of heat affected zones in the beltline region.
The applicant trust submit actual charpy V-notch impact test data from all beltline materials to demonstrate the upper shelf enargy of the weld metal, heat affected zones and base metals exceed 75 ft-lbs.
Comoliance with Accendix H,10 CFR part 50 The materials surie'.llance programs at Comanche Peak Unit 1 and 2 will be used to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region, resulting from exposure to neutren irradiation and the thermal environment. Under Cccanche Peaks' surveillance programs, fracture toughness data will be obtained from material specimens that are representative of the limiting base, weld, and heat-affected :ene materials in the beltline region. These data will permit the determination of
. the conditicas under which the vessel can te ;ers:ed with adecus:e argins of safety against fra::ure thr:ugnout its service life.
The fracture toughness properties of reactor vessel beltline caterials must be renitored :nreugncut the service life of C: ancne Peak Units 1 and 2 by I
a materials surveillance program tha: Tee:s :ne requirements of Asm Standard E 185-73, "5:andard Rec: : ence: Pratice for surveillance Tests for Nuclear Reacter Yessels" and A;;endix H of 10 CPR Part 50.
'Je have evaluated that apolicants's infor atien for degree of :::aliance to these requirements.
Eased en our evaluaticn we c:nclude :na: the a:;1ican:
has :et all the require ents of A;;endix H,10 CFR Part 50 wi:n :ne ex:e::ica of Paragra;n II.3. Paragra;h II.3 of A;;endix H requires :ta: :ne surveillance program c::aly with ASW E-135-73. AS W E-135-73 requires the surveillan:e ca;sule materials te removed frc= beltline reac r vessel case metals anc weld saccles whien represent the material that may 11:f: cpera:ica of :ne reac r vessel during its life: fee. The applican: has not iden:1fied fr =
wnich samples the material surveillance specimens were removed. To de cnstrate c:epliance with Paragraph II.3 of A;;endix H. :te a:alican: mus::
a) provide for eaca base.e:al and heat affected ::ne surveillance specimen the specimen ty;e, the orientatien of the specimen relative to the
- rinciple rolling direction of the pla
- e, : e nea: nwn er, :ne ::::enen c:de no. fr:m wnich the sa:cle was removed, enemical c: ;csition, es:ecially :nc a
the c:p;er (Cu) and ;nos;norus (P) ::ntents, the mei ing prac-ice anc the hea: treatmen received by the sar:le material.
- 3) provide f:r ea:n weld retal surveillance speciren :ne weid iden:ift:a:icn fr:a wnica :ne sarole was re cted, :ne weld wire :y;e anc nea; icen:ifica:i:n, flux ty;c and lo: icentifica:icn, wel: ;r: cess and nea: :rea::en: used
. for fabrication of the weld sacole, c) provide a sketch which indicates the a:imuthal location for each capsule relative to the reactor core.
ASTM E 185 requires the monitoring of tensile properties and impact properties to determine the extent of radiation induced change in mechanical properties.
/
Tensile properties have not been provided for any reactor vessel beltline materials. The applicant must provide tensile data for all weld _
,e metal and base metal used in the reactor vessel beltline.
Conclusions for Ocmoliance with Accendices G and H,10 CFR Part 50 Based on our evaluation of compliance with Appendices G and H,10 CFR Part 50, we conclude that the applicant nas not met all the fracture t:ughness require-ments of Appendix G and surveillance program requirements of Appendix H.
The areas of noncompliance include Paragraph I.A, I.C. III.B.4, IV.A.1, IV.A.3, and IV.A.4 and IV.S of Appendix G and Paragraph II.B of A:pendix H; these items will remain open in our safety evaluation until the applicant submits the necessary data.
Appendix G, " Protection Against Non-Ductile Failure,"Section III of the ASME Boiler and Pressure Vessel Code, will be used, together with the fracture tougnness test results required by Appendices G and H,10 CFR Par: 50, to calculate :he reactor coolant pressure boundary pressure-temperature limi-tations for the Cc=anche Peak Station.
The fracture toughness tests required by the ASME Coce and cy Ap;encix G of 10 CFR Part 50 will provide reasonable assurance tha adecuate safe:y =argins agains: t.te ;ossibility of ncn-ductile benavior or rapidly prc:aga:ing fracture can be establisted for all pressure re:af ning c :penents of :ne reac::r c:: an:
9 boundary. The use of Appendix G,Section III of :ne ASME Code, as a guice in establishing safe operating procedures, and use of :ne results of the i
fracture toughness tests performed in accordance with the ASME Coce and NRC regulations, will provide adequate safety =argins curing operating, testing, maintenance, and anticipated transient conditicns. Compliance with :nese Code provisions and NRC regulations ccnstitutes an acceptable basis for o
satisfying the fracture toughness requirements of General Cesign Critericn 31.
av The materials surveillance program, required by Accendix H,10 CFR Part 50, will provide infor=ation on material properties and the effects of irradia-tion on material properties so that cucr:ges in fracture toughness of material in Comanche Peak Units 1 and 2 reactor vessel beltlines caused by exposure to neutron radiation can be properly assessed, and adequate safety margins against the possibility of vessel failure can be provided.
Compliance with ASTM E-lSS-73 and Appendix H,10 CFR Part 50. assures that the surveillance program constitutes an acceptable basis for monitoring radiation induced changes in the fracture toughness of tne reactor vessel material and satisfies the materials surveillance requirements of General Design Criterion 31 and 32.
5.3.2 Pressure-Temoerature Limits Appendix G, " Fracture Toughness Requirements." and Appendix H. " Reactor Vessel Material Surveillance Progran Requirements," 10 CFR Part 50, describe the conditions that require pressure-tenperature limits for the reactor a
coolant pressure boundary and provide the general bases for these limits.
These appendices specifically require that pressure-temperature limits must provide safety margins for the reactor coolan: pressure boundary at least as great as the safety margins recommended in the ASME-Boiler and Pressure t
. Vessel Code,Section III, Appendix G, " Protection Against tion-Ductile Failure." Appendix G,10 CFR Part 50, requires additional safety margins whenever the reactor core is critical, except for low-level physics tests.
The following pressure-temperature limits imposed on the reactor coolant pressure boundary during operation and tests are reviewed to ensure that they provide adequate safety margin against non-ductile benavior or rapidly propagating failure of ferritic components as required by General Design Criterion 31:
1.
Preservice hydrostatic tests, 2.
Inservice leak and hydrostatic tests, 3.
Heatup and cooldown operations, and 4.
Core operation.
The applicant has indicated that technical specifications will be developed to establish pressure-temperature limits for normal operation and testing.
These limits are to be based on tiUREG-0452 Rev. 2 dated July 1979, " Standard Technical Specification for Westinghouse Pressurized Water Reactors", in accordance with the requirements of Appendix G 10 CFR Part 50 and the methodology of Appendix G Section III of the ASME Code. The applicant must provide the actual pressure-temperature limits for Comanche Peak Units 1 and 2 based upon the fracture toughness of the limiting reactor vessel material and predicted shift in the adjusted r2ference temperatures,RT
, resulting from radiation fiDT.
damage during 10 effective full power years. The pressure-te perature limits for the following conditions must be included in the technical specificatiens when they are submitted:
1.
Preservice hydrostatic test, 2.
Inservice leak and hydrostatic tests,
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re:hed:legy of Stancard Review Pian Se::icn 5.3.2 "Fressure-Te perture Li=its".
5.3.3 Remeter '.'essel In:e;rity We have revieaed ne felicein; F5AR se :i:ns reisted :: : e rea:::r vessel integri y for C::anche Feak Units 1 and 2.
A1:n c;n ecs: areas are revie-e:
se;arately in ace:r:ance with other review plans, rea:ur vessel i.:e::ri:v is of su:n ir::r:ance tha; a spe:ial surnary review cf all fa:::rs relating ::
rea:: r vessel integrity is warranted.
. 'de have reviewed the infomation in each area to ensure that it is complete and that no inconsistencies exist that would reduce the certainty of vessel integrity. The areas reviewed are:
1.
Cesign (SER 5 5.3.1) 2.
Materials of construction (SER 5 5.3.1) j 3.
Fabrication methods (SER $ 5.3.1) 4 Operating conditions (SER 5 5.3.2)
'de have reviewed the above factors contributing to the structural integrity of the reactor vessel and conclude that the applicant has complied witn Appendices G and H,10 CFR Part 50, except for the following items:
Paragraph III.B.4, Appendix G: The applicant has not provided sufficient infomation to detemihe whether individuals perfoming fracture toughness tests were qualified by training and experience and that the individuals had demonstrated competency to perfom tests in accordance with a written procedure.
Paragraph IV.A.1, Appendix G: The applicant has not provided sufficient information to. define the reference temperature, RT
, for all ferritic tiDT pressure retaining materials in the reactor pressure vessel.
Paragraph IV.B. Appendix G: The acolicant has not provided sufficient infnreation ::
define the upper shelf energy for all beltline materials.
Paragraph II.B. Appendix H: The surveillance capsule identification data per ASTM E-185 have not been included in FSAR.
In addition the applicant has not submitted pressure-temperature limits curves to ensure the safe operation of the reactor vessel during normal operaticn and testing.
. Until the applicant has supplied the information'necessary to cceplete our evaluation of compliance with Appendices G and H,10 C'FR Part 50 and reactor vessel pressure temperature limits, we can not complete our evaluation of the structural integrity of the reactor vessels of Comanche Peak Units I and 2.
. 4.1.1 Pumo Flywheel Intearity General Design Criterion 4, " Environmental and Missile Design Bases," of Appendix A,10 CFR Part 50, requires that nuclear power plant structures, systems and components important to safety be protected against the effects of missiles that might result from equipment failures.
Because flywheels have large masses and rotate at speeds of approximately 1200 revolutions per minute during normal operation, a loss of flywheel integrity could result in high energy missiles and excessive vibration of the reactor coolant pump assembly. The safety consequences could be significant because of possible damage to the reactor coolant system, the containment, or the engineered safety features. Adequate margins of safety and protection against the potential for damage from flywheel missiles can be achieved by the use of suitable material, adequate design, and inspection.
According to 5 5.4.1.5.2 of the FSAR, the materials used to manufacture the pump flywheels is SA-533 Grade B Class 1 steel plate. The applicant further states that the NDTT, nilductility transition temperature, of the flywheel material is no higher than t10*F, and that the Charpy upper shelf energy level in the
" weak" direction is no less than 50 ft-lbs at 70'F.
Minimum fracture toughness properties are ensured because the operstDg temperature of the flywheel is at least 100*F_ above the NDTT and 9e ena0c stress intensity factor, K 1/2 Id is in excess of 100.ksi-in.
Aw,m to demonstrate compliance with
1:
Safety Guide 14, October 27, 1971, Paragraph C.1 the applicant ust submit the Charpy V-notch impact and tensile property test data for eacn flywneel plate.
The pump flywneels are designed to the requirements of Paragrapn C.2 of Safety Guide 14 and the flywheel assembly are given a preope-'tional test I
at the design overspeed of the flywneel.
es-The applicant has stated that the inservice inspection program for the flywheel is defined in Paragraph 5.2.4 of FSAR.
Paragrapn 5.2.4 of FSAR is for inservice inspection and testing of reactor coolant pressure bouncary and does not include a section for inspection of the flywheel. The applicant must submit an inservice inspection program for the pump flywheels whicn complies with Paragraph C.4 of Safety Guido 14.
Until the applicant supplies the necessary information, we cannot conclude that the reactor coolant pump flywheels in Cccanche Peak Units 1 and 2 possess a margin of safety against flywheel missiles equivalent to that recommended in Safety Guide 14. Compliance with Safety Guide 14 will provice a basis acceptable to the staff for satisfying the requirements of General Design Criterion 4.
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. FRACTURE PREVENTION OF CONTAIimENT PRESSURE BOUNDARY Our safety evaluation review assest:- the ferritic materials in the nuclear plant containment system that constitute the centainrent pressure boundarv to determine if the material fracture toughness is in comoliance with the reauire-i o
ments of General Design Criterion 51, " Fracture Prevention of Containment r
Pressure Boundary".
GDC 51 requires that under operating, maintenance, testing and postulated accident conditions, (1) the ferritic materials of the containment pressure boundary behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.
1 The Comanche Peak Steam Electric Station containment is a reinforced concrete structure with a thin steel liner on the inside surface which serves a leaktight membrane. The ferritic materials of the containment pressure boundary, which are considered in our assessment, are those which have been applied in the fabrication of the equipment hatch, personnel lock, penetrations and piping system components including the valves required to isolate the system. These components are the parts of the ontainment system which are not backed by concrete and must sustain loads.
The Comanche Peak containment pressure bcundary is comprised of ASME Code Class 1,
< 2, and MC components.
In late 1979, we reviewed the fracture toughness requirements of the ferritic materials of Class MC, Class 2 and Class I components which
. typically constitute the containment pressure boundary. Based on this review we cetermined that the fracture toughness requirements contained in ASME Code Editions and Addenda tyoical of those used in the design of the Oceanche Peak
- containment may not ensure conpliance with GDC 51 for all areas of tne containmer.:
pressure boundary. We initiated a program to review fracture tougnness requirements for. containment pressure boundary materials for the purpose of defining those fracture toughness criteria that most appropriately address the requirements of GDC 51.
Prior to completion of this study, we have elected to apply in our licensing reviews, as an interim requirement, the criteria identified in the Sum er 1977 I
.Adden of Section III of the ASME Code for Class 2 components.
Because the criteria which have been applied in construction differ in Code classification and Code editions and addenda, we have chosen the criteria in the Sumer 1977 Addenda of Section III of the Code to provide a uniform review, consistent with the safety function of the contaic,ent pressure boundary materials.
The FSAR for Comanche Peak, however, does not provide the information necessary to characterize the fracture toughness of the materials of the reactor containment pressure boundary within the context of GDC 51. We request, therefore, the information requested in Question 123.10 be provided.
a.
ENCLOSURE 2 IE:UEST :CR CCIT*:NAL :NFCRMAT::N C'"A"C9E ?EAK STE18 ELECTR:C 37;7;CN. UN:T3 1 'E
- CCXET NCS: 50 ;;5 250 50-446 123.0 MATERIALS ENGINEERING BRANCH - CCF.PCNENT IYTEGRITY SECTION i
123.1 Identify whether SA-540 Class 1 or 2 material was used for closure boiting a~
in the reactor coolant pump.
If SA-540 Class 1 or 2 material was used for closure bolting in the rea;:or coolant pump, demonstrate the generic adequacy of the fracture toughness and demonstrate compliance with Paragraph I.C of Appendix G, to 10 CFR Part 50.
123.2 Indicate whether the individuals performing the fracture tougnness tests were-qualified by training and experience and whether their c:mpencency was demonstrated in accordance with a written procedure.
If the above information can not be provided, state why the information can nct be provided and identify why the method used for qualifying individuals is equivalent to those of Paragraph III.B.4 Appendix G,10 CFR Part 50.
l 123.3 Provide fracture toughness data for all ferritic materials used for l
l reactor coolant pressure boundary applications in vessels, piping, valves l
l and fasteners. The applicant has submitted some of the required data for reactor vessel materials. The data that remain to be submitted for vessel materials other than fasteners is identified in the attached table by tne designation " submit". Each Charpy V-notch impact test result must be l
reported. The 35 mils lateral expansion and 50 ft-lbs energy absorbtion level, if determined, must be submitted in a graphical form, with each data point plotted. The fracture toughness data submitted in Tables 5.3-2A, l
5.3-23, 5.3-7A and 5.3-75 in the FSAR should be consolidated with the data i
requested in the attached Table.
l
IAntt i ADillil0NAl. TRACIURC 10l!Gillit55 DAIA 10 B[ 50DHilltD DY liff APPt ICAllT TO Cattrtf Willi PARAGRAril IV.A.1 AND IV.D OF APPtreulX G LOCAlloit
' Drop Weight Charpy V-notch Charpy V-notch Reference Upper Shelf Upper Shelf NDI Iceperature long. oriented trans. oriented Tenperature Charpy V-notch Charpy V-notrh test results test results long. oriented trans. oriented i
test results test results Reactor Vessel Beltline H.sterial Weld Metal Sutani t Sulmit IIAl Sulal t Sutant t Sutal t Base Hetal Sutalt Sulal t Sutsen t t Suteal t All other Reactor Vessel Material WeId Hetal Sulant t Sulant t Sist eiel t 184 1 Sulmi t Sulalt N
Base Hetal Sulai t Pressurlier and Prinury $1de of 5 teams Generator Weld Metal Sulait Sutaal t Sutalt ilAl Sulal t Sulatt B.ise Metal Sulait sotal t suteal t l
q
3-If any of the above requested infermation cannot be provided, state why the informatica can not be provided, provide alternative test results for estimating the test data required by A;;endix G,10 CFR Part 50, and pro.*ide technical justification for the metheds used for estimating the required data.
123.4 Identify the location of each capsule and the materials in ea:n capsule.
t a) For each base metal and heat affected :ene surveillance specimen provide
<c-the specimen type, the orientatien of the specimen relative to tne principle rolling direction of the plate, the heat number, the cc=penent c de number from which tne sa pie was removed, the chemical cc ; sition especially the ce;;er (Cu) and pncsphorus (P) c:nten*s, the celting practice anc the heat treatment received by the sac;ie caterial, b) For ea:h weld me:al surveillance speci en provide tie wel identification frcm wnich the sample was rencved, the weld n' ire type anc heat icentifi-cation, flux type and lot identification, weld ;r: cess and heat treat-ment used f r fabrication of the wel: sa:;ie.
c) Frevide a sketch ahich indicates the a:itutnal iccation fer. eacn capsule relative to the reactor core.
123.5 Provide tensile data for all weld etal and base netal used in the reactor vessel beltline.
123.5 Frevica act.ai pressure-te :erature limits for C::an:ne Feak '.' nits I and 2 a
s
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.- a. #. =. - a...- a.. *. -. a..' *. ". r =.,.2 7 esu't..
e..
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ing fr:= ra:ir.ti:n damage during 10 affective fail : wer years. The pressure-
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s =.. '. '. '...=. '.... s a.. =. q.. =.. > 2.=. s.. - i... =.-
. 1.
Preservice hydrostatic tests 2.
Inservice leak and hydrostatic tests 3.
Heatup and cooldown operations, and 4.
Core operation 123.7 Provide fracture toughness and tensile property test data for each pump flywheel plate.
<r 123.8 Submit for review an inservice inspection program for the pump flywheels which complies with Paragraph C.4 of Safety Guide 14, October 27, 1971.
123.9 Revise the FSAR to indicate that the conclusions of Westinghouse Topical Report WCAP 9292 is applicable to Co:.anche Peak Units 1 and 2 SA-533 Grade A, Class 2 steel.
123.10 The FSAR for Comanche Peak does not provide the information necessary to characterize the fracture toughness of the materials of the reactor containment pressure boundary within the context of GDC 51. We request, therefore, the following information be provided the Materials Engineering Branch for review:
1.
Penetrations a.
Listing of all containment hot and cold pipe penetrations and related supplemental information which identifies penetration assembly sleeve, 4
process pipe and end closure materials by specification, final heat treat condition, nominal 00, and schedule, wall or section thickness, b.
Full size assembly detail drawings showings as built configurations and dimensioning of hot and cold pipe pentrations.
2.
Ecuicment Hatch a.
Full size assembly drawing and detail drawing of the hatch head assembly (barrel-bulkhead-door-interior).
. b.
Supplemental information identifying the materials of construction of the hatch head assembly by specification, final heat treat condition and section thickness.
3.
Personnel Access Lock y
a.
Full size assembly drawing and detail drawing of the door bulkhead as-assembly (barrel-bulkhead-door-interior) b.
Supplemental information identifying the materials of construction of the door bulkhead assembly by specification, final heat treat condition and section thickness.
4.
Main' Steal, Main Feedwater, Auxiliary Feedwater Systems a.
Full size piping diagrams and related supplemental pipe line lists and pipe line design / class specifications which identify.the systems of interest by line designators, pipe size and schedule, and pressure boundary materials specifications in addition to valve type, number and pressure boundary materials specifications.
b.
Piping diagram legend information.
We request that fracture toughness data be provided for the ferritic materials of those parts of the above components which, in the performance of the containment function under the conditions cited by GDC 51, provided a pressure boundary.
For those ferritic materials for which fracture toughness data are unavailable the a
following information is requested:
Seamless Pipe:
1.
Billet heating tengerature prior to piercing 2.
Intermediate reheat temperatures
. 3.
Stock wall thickness pricr to final sizing 4.
Reheating temperature prior to final sizing 5.
Pipe final heat treatments or pipe assembly heat treatnents Seamless Ells:
1 1.
Stock heating temperature prior to hot forming er 2.
In process reheat temperatures 3.
Ell final heat treatment or pipe assembly heat treatments.
Welded Pice:
1.
Metallurgical heat treat condition of plate stock as entered into
~
fabrication 2.
Plate stock heating temperatures prior to hot forming 3.
In process reheat temperatures 4.
Pipe final heat treatments or pipe assembly heat treatments.
Welded Elis:
1.
Metallurgical heat treat condition of stock as entered into fabrication 2.
Stock heating temperatures prior to hot forming 3.
In process reheat temperatures 4
In process heat temperatures 5.
Ell final heat treatment or pipe assembly heat treatments Valves:
a 1.
Final metallurgical heat treat condition of the materials of those valves parts which constitute parts of the pressure boundary
- 2. In-process post-weld re: air and intermediate heat treatments of the naterials of these valves parts which constiture parts of the pressure boundary.