ML17331A609
| ML17331A609 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/21/1981 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Dolan J INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| Shared Package | |
| ML17331A610 | List: |
| References | |
| NUDOCS 8102060548 | |
| Download: ML17331A609 (28) | |
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I: <<IT: 'gPji KL UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 0ockets Nos.
50-313 aod 0-31 January 21, 1981 Nr. John Dolan, Vice President Indiana and Michigan Electric Company Post Office Box 18 Bowling Green Station New York, 1Iew York 10004
Dear Mr. Dolan:
We have completed our review of the fracture toughness of the steam generator and reactor coolant pump supports at D.
C.
Cook Plant, Units Nos.
1 and 2.
The information concerning the support materials and their fracture toughness was ~upplied by Indiana and Michigan Electric in a letter dated November 23, 1977.
The information, initially reviewed by Sandia. Laboratories, was indepen-dently reviewed by Franklin Research Center.
Franklin reviewed the infor-mation per the criteria presented in NUREG-0577 and concluded the supports possess adequa e frac ure oughness to merit a Group III rating.
We agree with this finding and consider the steam generator and reactor coolant pump supports at D.
C.
Cook, Units 1
and 2 adequate with respect to frac-ture toughness and are therefore acceptable.
A copy of our Safety Evalua-tion and the Franklin Research Center Technical Evaluation Reports are enclosed for your information.
i cerely, ev at'g, C ief
Enclosures:
1.
Safety Evaluation 2.
FRC Tech.
Evaluation Report cc w/enclosures:
See next page Operating Reactor ranch nl Division of Licensing g(D2od,o& T(C
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,I 4 t,
Mr. John Dolan Indiana and Michigan Electric Company cc:
Mr. Robert W. Jurgensen Chief Nuclear Engineer American Electric Power Service Corporation 2 Broadway New York, New York 10004 Gerald Charnoff, Esquire
- Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.
Washington, D. C.
20036 Citizens for a Better Environment 59 East Yan Buren Street Chicago, Illinois 60605 Maude Preston Palenske Memori al Library 500 Market Street St. Joseph, Michigan 49085 Mr. D. Shaller, Plant Manager Donald C.
Cook Nuclear Plant P. 0.
Box 458 Bridgman, Michigan 49106 U. S. Nuclear Regulatory Commission Resident Inspectors Office 7700 Red Arrow Highway Stevensville, Michigan 49127 Mr.
Wade Schuler, Supervisor Lake Township
- Baroda, Michigan 49101 Mr. William R-Rustem (2)
Office of the Governor Room 1 - Capitol Building
- Lansing, Michigan 48913 Honorable James
- Bemenek, Mayor City of Bridgman, Michigan 49106 Director, Criteria and Standards Division Office of Radiation Programs (ANR-460)
U. S. Environmental Protection Agency Washington, D. C.
20460 U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN:
EIS COORDINATOR 230 South Dearborn Street Chicago, Illinois 60604 Maurice S. Reizen, M.D.
Director Department of Public Health P. 0.
Box 30035
- Lansing, Michigan 48909 William J., Scanlon, Esquire 2034 Pauline Bou'levard Ann Arbor, Michigan 48103
Enclosure 1
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 0
+~
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+~*g4 SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATIOH DONALD C.
COOK UNITS 1
AND 2 FRACTURE TOUGHI'IESS OF STEAM GENERATOR AHD REACTOR COO ANT P
MP SUPPORTS DOCKETS NOS.
50-315 5 316 1.0 Introduction 2.0 In a letter dated November 23,
- 1977, Indiana and Michigan Power Company responded to the NRC's request for information concerning the fracture toughness design of the steam generator and reactor coolant pump supports at D.C.
Cook Units 1 and 2.
This information was initially reviewed by Sandia Laboratories as part of generic technical activity A-12.
Sandia's evaluation concluded that the steam generator and reactor coolant pump supports at D.C.
Cook Units 1
and 2 were considered to be as good as careful, reasonable engineering practice can produce.
This evaluation is contained in NUREG-0577.
The information contained in the licensee's response was subsequently transmitted io Franklin Research Center for an independent evaluation.
This evaluation was based on the criteria presented in the draft version of NUREG-0577.
Evaluation 3.0 The enclosed Technical Evaluation Report was prepared by the Franklin Research Center as part of NRC Contract Ho. NRC-03-79-118.
The scope of Franklin's review was limited to those supports originally reviewed by Sandia which included the steam generator and reactor coolant pump supports.
Subsequent to the initial review by Sandia Laboratories, the scooe of Generic Technical Activity A-12 has been expanded to include other supports.
For those supports identified by the final version of HUREG-0577 that are not included in this review, the licensee 'should follow the implementation plan specified in NUREG-0577.
Conclusion Based on our review of Franklin's Technical Evaluation Report, we agree with their conclusion that the steam generator and reactor coolant pump supports at D.C.
Cook Units 1
and 2 possess adequate fracture toughness to merit a group III rating per. NUREG-0577.
Therefore, we consider the fracture toughness of the steam generator and reactor coolant pump supports acceptable.
Fg
Enclosure 2
TECHNfCAL EYAL'JATlON REPORT FR4CTURE YOUG HNESS OF STEAM G ENERATOR AND REACTOR COOt aNT pUMP SUPPORTS INDIANA ~ HICHI GA"~ Po~;R CONPAN DONALD C, COOK NUCLEAR PORER PUNT UNITS 1 AND 2 NRCDOGKETNO.
50 3)j and
$0 316 NRC TAG NO.
08479 and 08486 NRC CONTRACT HO. NRC43-79-'Its PRC PROJECT CS257 PRCTASK
'67 and 168 Prepared by Franklin Research Center The Parkway at Twentieth Street Philadelphia, PA 19163 Authors:
T.C.Scilvell, A.G.AI1cen, K.E. Do".schu, P.N. Noell FRC GrouP Leader:
T. C.ScQ.ue11 PfePared fof Nuclear Regulatory Commission Washington, D. C, %555 Lead NRC Engineer:
J. R.Fair Revision 1, November 1980 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or Implied, or assumes any legal liability or responsibility for any third party's use, or the results of such
- use, of any information, apparatus, product or process disclosed ln this report, or represents that its use by such third party would not infringe privately owned rights, IU Franklin Research Center iilljj A Division of The Franklin institute
~he ~jane Kronur9~. ptw'e.. pe. l9INQls) 444. inc
TERW5257-167/168 (Rev.
1)
CONTENTS Section 1
SUMMARy Title Pncne 2
INTRODUCTION 3
BACKGROUND
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4 CRITERIA APPLIED IN THE EVALUATION 4.1 Fracture-Toughness Grouping of Materials Used in Support Construction 4.1.1 Criterion 4.1. 2 Interpretation.
4 4.2 Plant Grouping for Fracture-Toughness Ranking of S/G and RCP Support Structures 4.2.1 Criterion 4.2.2 Znterpretation.
4.3 Criteria for Fracture-Toughness Adequacy of S/G and RCP Supports 4'.1 4'.2 NDT Criteria for Screening.
Interpretation.
4.3.3 Alternative Criteria 5
TECHNICAL EVALUATION 5.1 Review Procedure and Implementation of NRC Criteria 7
7 5.2 Extent of FRC's Review.
10 5.3 Review Findings 10 5.3.1 Use of Group I Materials in Applications Important to Structural Integrity of Supports 5.3. 2 Thick Section Use of Group II Materials in Applications Important to Structural Integrity.
.10 10
TERW5257-167/168 (Rev.
1) 5.3.3 Thin Section Use of Group II Materials in Applications Important To Structural Integrity 0
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5.3.4 Use of Materials Classified Group III by NUREG 0577, Upon Condition.
5.3.5 Use of Materials Classified Group III by NUREG 0577, Outright 6
CONCLUSIONS 10 ll 12 mmmm TABLE Title Pacae 5 1 COMPONENT SUPPORT
SUMMARY
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I)l) Franklin Research Center hbMaon ol The Fnnl4n~e iv
TER-C5257-167/168 (Rev.
1) 1
SUMMARY
Information concerning aspects of the fracture-toughness design of the steam generator (S/G) and reactor coolant pump (RCP) supports for the Donald C.
Cook Nuclear Power Plant Units 1 and 2 was submitted to the Acting Director of the Office of Nuclear Regulation by the Indiana and Michigan Power Company (IMPC) by letter dated Nov. 23, 1977.
This information was reviewed at the Franklin Research Center (FRC) and evaluated in accordance with the criteria of the Nuclear Regulatory Commission (NRC) as set forth in NUREG 0577-Draft (henceforth referred to simply as NUREG 0577).
The information had previously been reviewed as part of the preparation of NUREG 0577<
and D. C.
Cook Units 1 and 2 had been assigned a Group III (rela-tively best) plant ranking for fracture toughness of S/G and RCP supports.
This ranking was regarded as tentative.
Subsequently, the NRC requested FRC to conduct an independent review prior to finalizing the ranking.
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WC's review was confined to fracture-toughness issues in supports above the embedment.
The review was conducted in accordance with NRC criteria and to a procedure standardized for the several licensees whose support designs were reviewed at FRC.
As a result of its review, FRC confirmed that the Group III plant ranking assigned to Donald C.
Cook Nuclear Power Plants Units 1 and 2 for fracture toughness of S/G and RCP supports is justifiable.
2 ~
INTRODUCTION l g This report provides a technical evaluation of information supplied by IMPC with its letter of Nov. 23, 1977, to Mr. Edson G. Case, Acting Director Office of Nuclear Regulation.
The information concerns the fracture-toughness design of supports for the S/Gs and RCPs for D. C.
Cook Units 1 and 2.
The objective of the evaluation is to rank the design for fracture-toughness integrity on a relative scale in accordance with the grouping scheme and criteria established in NUREG 0577.
(l1 Franklin Research Center A~ ot The Fnnkhn~
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TER&5257-167/168 (Rev. 1) 3 ~
BACKGROUND During the course of the NRC licensing review for -two pressurized water reactors (PHR) r North Anna Units 1 and 2, questions were raised regarding the fracture-toughness adequacy of certain members of the S/G and RCP supports.
The potential for lamellar tearing in some support members was also questioned.
The staff's concern in the North Anna licensing process was that perhaps
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not enough attention had been given to the selection of materials for, and fabrication of, the S/G and RCP supports.
Fracture toughness of a material is a measure of its capability to absorb energy without failure or damage.
Generally, a material is considered "tough" when>
under stated conditions of stress and temperature, the material can withstand loading to its design limit in the presence of flaws.
Toughness also implies that, under certain conditions, the material has the capability to arrest the growth of a flaw.
A lack of adequate toughness (accompanied by the combination of low operating temperature<
presence of flawsr and nonredun-dancy of critical support members) could result in failure of the support structure under postulated accident conditions, specifically a lossmfmoolant accident (LOCA) and safe shutdown earthquake (SSE)
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To address fracture-toughness concerns at the North Anna facility, the licensee undertook tests not originally specified and not included in the relevant ASSN specifications.
These tests indicated that material used in certain support members had relatively poor fracture toughness at 80'F metal temperature.
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'n this case, the licensee agreed to raise (by ancillary electrical heat) the temperature of the S/G support beams in question to a minimum of 2254F every time, throughout the life of the plant> that the reactor coolant system (RCS) is pressurized above 1,000 psig.
The NRC staff found this to be an acceptable resolution.
Because similar materials and designs were used in other plants and be-cause similar problems were therefore possible, this matter was incorporated into the NRC Program for Resolution of Generic Issues as 'Generic Technical lllltl Franklin Research Center ADiiaen of The Fran IneOMt
TER-C5257-167/168 (Rev.
1)
Activity A 12 Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports."
Since the original licensing action (North Anna Units 1 and
- 2) involved only the S/G and RCP supports of PWRs, the staff's initial efforts were di-rected toward examination of the corresponding supports at other PWR facili-ties.
- However, the staff has kept in mind the possibility of expanding its review to include other support structures in PWR plants and support struc-tures in boiling wat'er reactor (BWR) plants.
The integrity of support embedments was not questioned during the North Anna licensing action; consequently, emphasis was placed on resolving the most immediate generic issue whether or not problems similar to those uncovered at North Anna exist at other facilities. It was the staff's judgment that inclusion of an evaluation of support embedments in the initial review would require detailed, plant-specific investigations that were beyond the scope of the preliminary, overall generic review.
Such considerations were deemed more suitecf to a subsequent phase when more detailed investigations of individual plants might be undertaken.
Requests for information were sent to licensees in late 1977r responses to these requests were received during 1978.
Sandia Laboratories in Albuquerque, New, Mexico, was retained to assist the staff in the review and analysis of the information received from licensees and applicants.
Based on analysis of this information< the technical studies performed by Sandia Laboratories, and review of the issues by the NRC staff, the NRC developed an NRC staff technical position on these issues<
which is presented in NUREG 0577r "Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports."
In addition, NUREG 0577 establishes criteria for evaluation of the fracture-toughness adequacy of S/G and RCP supports.
NUREG 0577 also applies h
certain of these criteria to the support structures of a number of PWR plants to achieve plant groupings according to the relative fracture-toughness inte-grity of these supports.
09 Franklin Research Center Abhsion d Tlk FgaHhn lemur M3w
TER-C5257-167/168 (Rev.
1)
The plant ratings are:
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Group I (lowest) e Group ZI (intermediate)
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Group III (highest)
During the generic study, a number of PWR plants were reviewed for the fracture-toughness adequacy of their RCP and S/G designs.
As a result of these
- reviews, each plant was assigned a tentative plant ranking of either Group I, IZ, or III.
Several Plantsg D
C Cook Units 1 and 2 among
- them, were tentatively ranked Group III.
In the appendix to NUREG 0577 prepared by Sandia Labora-
- tories, who initially established the rank'ngs which subsequently received NRC staff endorsement, the significance of the Group IIZ ranking is described as:
"considered to be as good as careful, reasonable engineering practice can produce."
Howevers before finalizing the tentative Group III rankings, the NRC requested FRC to conduct an independent review of the Group III plants (in s
conjunction with similar FRC task assignments to review the fracture-toughness adequacy of corresponding supports in certain other plants) and to prepare a
Technical Evaluation Report for each plant, presenting the review findings.
The technical evaluation reported herein applies the criteria of NUREG 0577 to the S/G and RCP supports for D. C.
Cook Units 1 and 2 to provide an assessment of the fracture-toughness adequacy of these supports leading to a plant ranking.
4 ~
CRITERIA APPLIED IN THE EVALUATION 4 ~ 1 FRACTURE-K)UGHNESS GROUPING OF MATERIALS USED IN SUPPORT CONSTRUCTION 4.1.1 Criterion Table 4.6, Material Groups, of Appendix C to NUREG 0577 groups materials according to their relative fracture toughness as:
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Group I (poorest)
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Group II (intermediate)
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Group ZII (best)
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'Jt!.'ranMln Research Center A Deduce d Tht Fnnkfin inseutc
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TER-C5257-167/168 (Rev.
1) 4.1.2 Interpretation If no supplementary requirements were called out in the material spe'cifi-cation aimed at procuring a product with fracture-toughness properties supe-rior to those routinely supplied under the material specification, then the material was grouped in accordance with Table 4.6.
If addit:ional requirements aimed at procuring a product with superior fracture<<toughness properties were specified, consideration was given to cred-iting this specific material order with an improved material-group rating.
4.2 PLANT GROUPING FOR FRACTURE-WUGHNESS RANKING OF S/G AND RCP SUPPORT STRUCTURES 4.2.1 Criterion Plants are classified on the basis of the construction materials used in the supports afte'r giving consideration to the importance of their location and function within the structure, and their consequent importance to support-structQre integrity.
(Refer to pages 5 and.6 of NUREG 0577, Part I.)
4.2.2 Interpretation Plants were assigned a plant-group ranking identical to the material-group ranking of the least fracture-tough material used in the construction, pro-vided this usage is important to support integrity.
4+3 CRITERIA FOR FRACTURE-TOUGHNESS ADE9UACY OF S/G AND RCP SUPPORTS It is the clear intent of NUREG 0577 that licensees demonstrate the fracture-toughness adequacy of the S/G and RCP supports or that they take I
appropriate corrective measures to assure their fracture-toughness integrity.
NUREG 0577 provides guidance for such demonstrations.
4.3. 1 NDT Criteria for Screening 30'Z NDT + 1.3tr +
or
( F)
- upports 60'F l)lj Franklin R',esearch Center A Divison d The f rain hsecutc
0
TER-C5257-167/168 (Rev.
1) where:
o NDT is the mean nil ductility transition temperature appro-'riate to the material as given by Table 4.4 of Appendix C
to NUREG 0577 o
a is the standard deviation for the data used to determine NUT as listed in Table 4.4.
o Tsupports is the lowest metal temPerature that the suPPort member will ever experience throughout the plant life when the plant is in an operational state.
In the absence of
- measured, plant-specific data, Tsuoports is taken as 75'F.
o The temperature term, 30'F or 60'F, is an allowance for sec-tion size (30 F for thin sections and 60'F for thick sec-tions).
4.3.2 Interpretation If evidence is furnished by the licensee proving that other values of NDTg
~1a, or g are actually valid for the S/G or RCp supports and materi-supports als in the licensee's
- plant, such data may be used.
.If acceptable alternative evidence is not available, the above-stipulated values should be used.
4.3.3 Alternative Criteria NUREG 0577 also recognized that fracture-toughness integrity is a complex matter involving a number of interrelated factors, most of which are plant specific.
Consequently, demonstration of compliance with the screening crite-
" ria is but one means of providing satisfactory assurance of fracture-toughness adequacy.
h NUREG 0577 not only recognizes that other means of showing compliance with the intent of NUREG 0577 are possible, but also offers extensive guidance re-lating to several approaches by which such a demonstration may be achieved.
Because of the plant-specific character that such demonstrations must take, NUREG 0577 does not restrict the licensees to any single approach but, instead, encourages each licensee to review the fracture-toughness adequacy of his S/G and RCP supports and submit evidence of his findings.
l)ll FranMin Research Center h DiKace et Th~'FnnkSn )needle
TER-C5257-167/168 (Rev.
1) 5 ~
TECHNICAL EVALUATION The information furnished to the NRC regarding the fracture toughness of, and the potential for lamel.lar tearing in, S/G and RCP supports at D. C.
Cook Units 1 and 2 was reviewed at FRC.
This information was supplied in response to the NRC staff's generic letter to PRR licensees concerning these issues.
A copy of the staff's request-for-information letter (in generic form) may be found in NUREG 05771 Appendix B.
Only fracture toughness issues were addressed in the FRC review; the review procedure is described below.
5 ~ 1 REVIEW PROCEDURE AND IMPLEMENTATION OF NRC CRITERIA The drawings and information submitted were first examined to become
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familiar with the structural design, material selection, and construction practices.
Key items from this information were condensed to tabular form and are presented in Table 5.1.
Zn accordance with a review procedure standardized for the licensees whose plants were evaluated at FRC, the first step was to compile a list of materials used in all members significant to the structural integrity of the S/G and RCP supports.
The listed materials were taken from these reported in the response to Item 1 of the NRC's request for information> supplemented by a survey of the support drawings for additional materials which might be indi-cated there.
To each of the materials so identified, two criteria tests were applied:
1.
The NDT criteria for screening (paragraph 4;3.1 of this report).
2.
The material group ranking in accordance with the procedures of Section 4.1.
For plants which used them, materials with an assigned Group I or Group ZI fracture-toughness rating were further categorized as thick or thin using the formula shown on the following page to determine the section thickness above which brittle (plain strain) behavior may be anticipated under dynamic load.
l 0
OTILlTT indiana 6 NlchlSan rover IIITLUILLS HSSS Mestinkbouse SUnt TASI.K S ~I COHPONENT SUPPORT SUNHARZ FIAHft Donald C Cook I 4 AK American Electric paver Company UU~ PSST
~ UIPLISS NAXINUH ALLOMhtkk DKSICN Stkt5$
TZPE Construct ioo Nster islet NILL CtkTS.
HEAT AVAIIAKLK TRKATNtNT NDK ON HATERIAL rkhcTURR TOUCNNK$5 TSUP HKHSRANK 6 RKNDINC (HORH4L) lllkO(jCH IlllCSIICSS A-615 Cr I A-36 A-555 Soltlnk Hater( ~ 1st A-I&3 SI A-IV4 Cr I A(51 4145 A 490 AISI 4340 Meldin5 Haterisl ~ t K6UXX~
E IUXX SOIa-CI ~ 5015-Cl ~ SOIR-O 5016-CI, SOI R&2 ~ 2-I/1E or 3-1/II Ni Content sub arc consumablea
'Tea Zes A-36 to f(ne-Sra(n practice.
Hormallaed A-355 ln Critical members.
UT under veld Tbru TITIcbneas
~reaa Reduced Ares Tests CVN for A-615 ~ A-36 ~
A-555 ((5 (t-lba 530 r).
Also ItAK and Meld Nsterlal ~
Norwi Upsets RISC HssTusl hllovablee Emerkencyi 0 5 5 Fsultekr Non-Linear Elastic-plastlc Anal ye I~
OU6$ Sy rhRRIChTION MKLDIHC FRUCkSS Hsnual Natal hrc Sub erc MELDIHC PROctDURE AISC Codes Se<<tion lk qual(fied pro-cedures tOST MELD!HO TREATNKNT Stress Relief NKTNOOS USED TO FRKVKNT LANKLLAR TEARINC A15C Code Joints HDR AND INSPECTIONS ptkroRNZ.D Mr or ST+ere
- possible, HF or LP QkSICN ZZPK OF SUPPORT rinuolumn CODt USED LOADIHC CONDITION5 Normal:
Dl. s TL Upset t DL s TL + Okt Esserkencyt DL s TL s 055 Faulted:
DL s 'TL s Dbts tk HINIHUH TKHPKRATURE OF SUPPORT 60'r (hmhlent temperature near supports)
TER-C5257-167/168 (Rev.
1)
The critical thickness is given by:
KIO 2
2.5 f
J tryD where:
syD is the dynamic yield strength of the steel.
KID is the nominal, minimum assured fracture toughness of the steel in accordance with values supplied by gUREG 0577.
tc is the critical thickness.
In members thicker than tc, brittle (i.e., plane strain) behavior may be expected.
A similar categorization for Croup IIImaterials was not deemed necessary for purposes of the review because such materials are sanctioned for thick-section use by virtue of their group rating.
Structural drawings were then examined for;
- l. All structurally significant uses of t;roup I materials.
2.
All structurally significant uses of Group II materials in thick sections.
3.
Structurally significant applications of materials known to be sensitive to stress corrosion cracking or other special failure mechanisms which might make them prone to brittle behavior.
The circumstances associated with such usage were then examined.
Consider-ation was given to factors such as:
direction of loadings,(always compressive or sometimes tensile), stress levels in the member as indicated in the licensee's
- response, the presence of stress raisers in member geometries, re-I dundancy of load paths, and the like.
Applications judged to be of problematic fracture toughness were identified for more detailed evaluation at a future date.
In addition, information furnished on welding and on material specifica-tions was examined for its fracture-toughness implications by a welding engi-neer and a metallurgist, respectively.
00 Franklin Research Center A~ ofTli~ FcankLn InsOaae
4
TER-C5257-167/168 (Rev.
1)
As a result of the review findings and in accordance with the criteria procedure described in Section 4.2 of this report>
a tentative plant ranking for fracture-toughness adequacy of S/G and RCP supports was assigned.
5.2 EXTENT OF FRC REVIEW FRC's evaluations were restricted to assessments of the fracture toughness of supports for steam generators and reactor coolant pumps.
Assessment of the fracture-toughness adequacy of supports for other components and of the embed-ment was not included in the scope of FRC's work assignment and was not inves-tigated.
5 ~ 3 REVIEW FINDINGS 5.3.1 Use of Group I Materials in Applications Important to Structural Integrity of Supports None found.
I 5.3.2 Thick Section Use of Group II Materials in Applications Zmportant to Structural Integrity None found.
5.3.3 Thin Section Use of Group II Materials in Applications Important To Structural Integrity AS'IM A-618 steel is indicated on both S/G and RCP support drawings as the material for the main vertical columns.
These are. constructed of 12 inch dia-meter, double-extra-strong pipe (i.e., seamless tube of 12 3/4 inch o.d.
and with 1 inch walls).
NUREG 0577 classifies ASTM A-618 as a Group ZZ steel when furnished as formed and without additional specification requirements.
- However, AS'IM A-618 Grade 2 was specified for this tubing and Charpy V-notch testing was required.
Specification ASTM A-618 Grade 2 limits silicon content to a maximum of 0.30 percent, and requires addition of vanadium.
The actual steel used was analyzed to have only 0.19 percent silicon (sufficient to completely deoxide 9l) Franklin Research Center A~ d Tbc Fnnkln hseua
e TER-C5257-167/168 (Rev.
1) the steel according to silicon-killed practice) and to contain 0.04 percent vanadium (which would tend to promote a finer grain size).
The test report furnished in the information supplied to NRC by IMPC indicated that the steel possessed a Charpy V-notch impact energy of 24 ft-lbs at 30'F.
This value, if typical of all heats, qualifies this steel to be of adequate quality and toughness for 1 inch section usage.
5.3.4 Use of Materials Classified Group IIZ by NUREG 0577, Upon Condition ASTM A-588 is the major component steel of both the S/G and RCP supports and was supplied as A-588, Grade A.
This steel is classified in NUREG 0577 as a Group ZI material in the as-rolled or hot-worked condition.
However< in sections 1/2 inch thick and over, the steel was ordered normalized and Charpy V-notch impact tests were required.
The test data furnished for review indi-r cate adequate toughness at 30'F in all thicknesses.
In view of the additional requirements specified, the A-588 steel used in this application is deemed to be of-sufficient quality and toughness to merit a Group III material rating.
C The text and materials table of the ZMPC letter of response refer to use of AS'IN A-36 steel as a material of construction for S/G and RCP supports in the Cook plants.
These also state that it was ordered to fine grain practice and required to be subjected to Charpy impact testing.
With such additional requirements the A-36 steel would be considered<
under NUREG 0577 criteria, as sanctioned for general use in S/G and RCP supports.
- However, FRC did not find it indicated as a material of construction on any of the drawings furnished for review nor could mill test or other material. data for this steel be found among the extensive information supplied.
Thus, although the reason for listing A-36 steel as a material of construction is not clear>
when oidered to the stated requirements A-36 is a Group ZZI steel, and its use would not affect the final classification of this plant.
5.3.5 Use of Materials Classified Group ZZZ by NUREG 0577, Outright All bolting and welding materials.
ill! Franklin Research Center Acevuion 8 tee Franklin Inane
TER-C5257-167/168 (Rev.
1) 6 ~
CONCLUSIONS The design and construction of supports for steam generators and reactor coolant pumps at Donald C.
Cook Nuclear Power Plant Units 1 and 2 have been reviewed for fracture-toughness adequacy at the FRC.
Criteria for the suitability of materials and construction practices for S/G and RCP supports were provided by the NRC staff as published in NUREG 0577-
/
Draft.
Zn the'eview, general criteria of NUREG 0577 were specifically applied to information furnished by Indiana and Michigan Po~er Company (ZMPC) concern-ing the supports in D.
C.
Cook Units 1 and 2.
The review was restricted to supports (above the embedment) for steam generators and reactor coolant pumps.
Conclusions relating to them do not necessarily extend to the support design of other components.
Zn the case of D. C.
Cook Units 1 and 2 FRC concludes that:
~t J l.
Engineering measures taken in support design, material selectioni material specification, material acceptance testing, fabrication
- methods, and inspections provide reasonable evidence that the steam generator support structures possess adequate fracture toughness to meet NRC criteria for a Group IIZ rating.
2.
Engineering measures taken in the design and construction of the reactor coolant pump supports provide similar evidence to qualify them for a Group III rating also.
3.
The Group ZIZ (relatively highest) plant rating for fracture-toughness adequacy of supports assigned to Donald C.
Cook Nuclear Power Plant Units 1 and 2 in NUREG 0577-Draft is justifiable.