ML17331A465
| ML17331A465 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/30/1980 |
| From: | Carfagno S FRANKLIN INSTITUTE |
| To: | Butcher E Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17326A752 | List: |
| References | |
| CON-NRC-03-79-118, CON-NRC-3-79-118 NUDOCS 8010080417 | |
| Download: ML17331A465 (22) | |
Text
REGULAT~ INFORMATION D I STR I BUT ION~i'STEM (RIDS)* 't ACCESSION NBR: 8010080417 DOC ~ DATE; 80/09/30 NOTARIZED: NO DO FACIL:50 315 Donald C, Cook Nuclear Power Plantg Unit lr Indiana S 500031 50 316 Donald C. Cook Nuclear BYNAME Power Plantg Unit 2g Indiana 8 05 6 AUTH AUTHOR AFF ILI AT ION CARFAGRO~S.P. Franklin Institute/Frankl.in Research .Center RECIP ~ NAME RECIPIENT AFFILIATION BUTCHER~K,J ~ Assistant'irector for Plant Systems
SUBJECT:
Forwards "Fracture Toughness of Steam Generator 8 Reacto,r Coolant Pump Suppo,rts." Rept descr,ibes criteria for 'reviewi.
procedure L basis for conclusion supporting NUREG 0577 Gt oup III plant ranking for fracture toughness adequacy.'ISTRIBUTION CODE: X004S COPIES RECEIVED! L(TR ENCL 1. SIZE!
TITLE: Frankl in Research Center Contract Repor t NOTES: ILE;3 copies all material 05000315>>
05000316 RECIPIENT COPIES RECIPIENT COPIKS ID COOK/NAME LITTR ENCL ID CODE/NA>>MK LTTR ENCL':
INTERNAL: A/D MATLl L QUAL 1 ~ 1 A/D SFTY ASSESS 1 1 BECKHITH C 1 1 BUTCHER e E"i. 15- 3 CONT SYS A 1 0 DIRg HUM FACt SFY 1 0 D IR g L'I CENS ING03 1 1 UAL~ BR 1 0 GLAGOI.A A 1- 1 REG F 01 1 1 OC p 9 jggp TOTAL'UMBER OF>> COPIES REQUIRED! LTTR ENCL<
~
t 1 A
I ~ 'll 1
r
'I "i ,0 A,
A 1 P ,
a . i 4
~ v A ii i AA f A'A A
t llf i r v
a s>
A A I
(l( Franklin Research Center A Division of The Franklin institute Sep'tember 30, 1980 I'
United States Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Edward J. Butcher, Jr .
Project Officer
Reference:
FRC Project C5257 NRC Contract NRC-03-79-118 NRC TAC No. 08479 and 08486
,FRC Task No. 167 and 168
Title:
FRC TER: Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports - D.C. Cook Units 1 and 2
Dear Mr.,
Butcher:
Enclosed is a Technical Evaluation Report which addresses the fracture-toughness adequacy of steam generator and reactor coolant pump supports in D.C. Cook Units 1 and 2.
The report describes the criteria established by NRC for this review, the review procedure used to evaluate plant compliance with the criteria, and the basis for FRC's conclusion supporting a NUREG 0577 Group III (relatively superior) plant ranking for fracture-toughness adequacy of these support s tructures ~
Very truly yours, S. P. Carfag o Project Manager
'SPC/mh j Enclosure 0 q cc: J. R. Fair (also K. R. Wichman reproducible copy) $
A. F. Glagola (letter only)
SIS g) 80ZO080 4 Z t g )(tn The Benjamin Franklin Parkway, Philadelphia. Pa. 19103 (215) 448-1000 TWX-710 670 1889
t 0 T
V no =! v oa 'one>f:n . 1 lg
'lA'i l
l A
/1t k
P
) 8 P ~ t d lt 5< ~ lf'l'p 4 t 4 3 V R ls'
0 TECHNlCAL EVALUATION REPORT FRACTURE TOUGHNESS OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS INDIANA R I'1ICHIGAN POWER .CONPANY .
DONAl D C COOK NUCLEAR POWER PLANT UNITS 1'ND 2
'NRC DOCKET NO. 50-315 and 50-316 NRCTACNO. 08479 and 08486 FRC PROJECT C5257 NRC CONTRACT NO. NRC43-79-118 FRCTASK 167 and 168 Prepared by Franklin Research Center Authors: T.C. S tilwell, A.G.Allten, The Parkway at Twentieth Street K;E.Dorschu, P.N.Noell Philadelphia, PA 19103 FRC Group Leader: T.C. S ti.lwell Prepared for Nuclear'Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: J.R.Pair September, 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 in this report, or represents that its use by such third party would not infringe privately owned rights.
IIII Franklin Research Center A Division of The Franklin Institute The Benjamin FranMin Pa~ay. Phiia.. Po. ) 9)03 Q)5) 448 ) 000
l~ '
TER-C5257-16 7/168 CONTENTS Section Title ~Pa e 1
SUMMARY
~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
2 INTRODUCTION 3 BACKGROUND ~ 2 4 CRITERIA APPLIED IN THE EVALUATION . ~ 4 4.1 Fracture-Toughness Grouping of Materials Used in Support Construction . 4 4.1.1 Criterion 4 4.1.2 Interpretation. 5 4.2 Plant Grouping for Fracture-Toughness Ranking of S/G and RCP Support Structures 5 4.2.1 Criterion 5 4.2.2 Interpretation. 5 4.3 Criteria for Fracture-Toughness Adequacy of S/G and RCP Supports 5 4.3.1 NDT Criteria for Screening. 5 4.3.2 Interpretation. 6 4.3.3 Alternative Criteria 6 5 TECHNICAL EVALUATION ~ 7 5.1 Review Procedure and Implementation of NRC Criteria ~ 7 5.2 Review Findings 10 5.2.1 Use of Group I Materials in Applications Important to Structural Integrity of Supports 10 5.2.2 Thick Section Use of Group II Materials in Applications Important to Structural Franklin Integrity . 10 5.2.3 Thin Section Use of Group II Materials in Applications Important To Structural Integrity . 10 00 Franklin Research center A !Xvldon ot The Instate
TER-C5257-167/168 5.2e4 Use of Materials Classified Group II by NUREG 0577, Upon Condition.
5.2.5 Use of Materials Classified .Group III by NUREG 0577, Outright 5.2.6 Issues Not Completely Resolved. 11 6 CONCLUSIONS 11 TABLE Number Title Page 5.1 COMPONENT SUPPORT
SUMMARY
. . . . . . . . . 8 (Ilf Franklin Research Center A Dtviston ot The Fronton Institute
TER-C5257-167/168 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.
FRC'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 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.
(ill Franjdln Research Center A Divisive of The Frsnidin Insotuie
TER-C5257-167/168
- 3. BACKGROUND During the course of the NRC licensing review for two pressurized water reactors (PWR), 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 teax'ing in some support members was also questioned.
The sta'ff's concern in the North Anna licensing process was that perhaps 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 flaws, and nonredun-dancy of critical support members} could result in failure of the support structure under postulated accident conditions, specifically a loss-of-coolant accident (LOCA) and safe shutdown earthquake (SSE).
To address fracture-toughness concerns at the North Anna facility, the licensee undertook tests not oxiginally specified and not included in the relevant ASTM specifications. These tests indicated that material -used in certain support members had relatively poor fracture toughness at 80'F metal temperature.
In 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 225'F 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 t)ll Franklin Research Center A OMshn of Ibe FrankKn batiste
TER-C5257-167/168 Activity A12 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 water 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 suited to a subsequent phase when more detailed investigations of individual plants might be undertaken.
Requests for information were sent to licensees in late 1977; responses to these requests were received during 1978. I 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 an analysis 'of the information, the technical stud-ies 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 0577, "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 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.
3 00 Franklin Research Center A bhteian at The FtenMin Ineetute
TER-C5257-167/168 The plant ratings are:
~ Group I (lowest)
~ Group II (intermediate)
~ 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, II, or III.
Several Plants, 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 rankings which subsequently received NRC staff endorsement, the significance of the Group III ranking is described as:
"considered to be as good as careful, reasonable engineering practice can produce."
However, before finalizing the tentative Group III rankings, the NRC reque'sted FRC to conduct an independent review of the Group III plants (in 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.
i 4~ CRITERIA APPLIED IN THE EUALUATION 4.1 FRACTURE-TOUGHNESS 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:
~ Group I (poorest)
~ Group II (intermediate)
~ Group III (best)
(ill Franklin Research Center A Ohfsfon of The Ffenkffn Insofufe
TER-C5257-167/168 4.1.2 Interpretation If no supplementary requirements were called out in the material specifi-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 additional 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-TOUGHNESS 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 after giving consideration to the importance of their location and function within the structure, and their consequent importance to support-structure. 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.
I 4.3 CRITERIA FOR FRACTURE-TOUGHNESS ADEQUACY 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 appropriate corrective measures to assure their fracture-toughness integrity.
NUREG 0577 provides guidance for such demonstrations'.3.1 NDT Criteria for Screening 30 F DT+ . + o ~T,u t,'0'F I Franklin Research Center h DMsion of The FcanklW Insatute
0 l TER-C5257-167/168 where:
~ NDT is the mean nil ductility transition temperature appro-priate to the material as given by Table 4.4 of Appendix C to NUREG 0577 '
~ tr is the standard deviation for the data used to determine NDT as listed in Table 4.4.
~ Tsupports is the lowest metal temperatur'e that the support member will ever experience throughout the plant life when t'e plant is in an operational state. In the absence of measured, plant-specific data, Tsupports is taken as 75'F.
~ 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 NDT, tr, or T are actually valid for the SlG 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.
NUREG 0577 not only recognizes that other means of shoving 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 SFG and RCP supports and submit evidence of his. findings.
ll( Franklin Research Center A Oivislon or The FsanMfn Insiitute
TER-C5257-167/168
- 5. TECHNICAL EVALUATION The information furnished to the NRC regarding the fracture toughness of, and the potential for lamellar 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 PWR licensees concerning these issues. A copy of the staff's request-for-information letter (in generic form) may be found in NUREG 0577, 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 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.
In 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 those 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 vere 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 vith an assigned Group I or Group II 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.
I FranMin Research Center A Melon d The Franklin Institute
TABLE 5.1 COHPONENT SUPPORT SU)DIARY PLANT: Donald CD Cook I 6 2 UTI LITY NESS AE SUPPORT SUPPLIER Indiana & Michigan Pover Meetinghouse American Electric Pover Company HATEkIALS HAXIHUH ALLOWABLE DESICN STRESS FRACTURE HILL CERTS. HEAT NDE OH TOUGHNESS HEHBRANE 6 THROUGH TYPE AVAILABLE TREATMENT MATERIAL TEST BENDIHG (NORMAL) THICKNESS Construction Hacerials:
h-618 Cr 2 Yes A-36 to fine-grain UT under veld Thru-'Thickness Normal-Upset: 0.65 Sy A-36 Yes practice. areas Reduced Area Tests AISC Hanual Allovables A-588 Normalized h-588 in Emergency:
Bolting Materials: Critical members. CVN for h-618, A-3&, 0.9 S h-588 (15 ft-lbs Paul ceItl h-193 B7 830'F). Non-Linear h-194 Cr 7 Also HAE and Meld Elascic-Plastic AISI 4145 Haterial ~ Analysis A&90 AISI 4340 Melding Hacerials:
E60XX> E70XX 8016-01, 8018-01, 8018-G 8016-C2, 8018-02, 2-1/2Z or 3-1/2Z Ni Content sub arc consumables FABRICATION METHODS USED TO NDE AND MELDING MELDING POST-MELDING PREVENT LAHELLAR INSPECTIONS PkOCESS PROCEDURE TREATHENT TEARING PERFORHED Hanual Hetal hrc AISC Code, Stress Relief AISC Code Joints llT or RT vhere Sub arc Seccion IX possible Qualified Pro- HP or LP cedures DESIGN TYPE OF SUPPORT CODE USED LOADING CONDITIONS HINIHUH TEMPERATURE OF SUPPORT Pin&ulnas> Normal: DL + TL 60 F (Ambient temperature near Upse t: DL + TL + OBE su ppor ts )
Emergency: DL + TL + DBE Faulted: DL + TL + DBEs PR
TER-C5257-167/168
,The critical thickness is given by:
2.5 [
"ID fryD
]
2 where.'yD 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 NUREG 0577.
tc is the critical thickness. In members thicker than tc, brittle (i-e., plane strain) behavior may be expected.
A similar categorization for Group III materials 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:
1 ~ All structurally significant uses of Group I materials.
- 2. All'tructurally 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-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.
9tj Franklin Research Center h Onfafon of 'aha Frrrnwrn Inrrfrrrre
TER-C5257-167/168 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 REVIEW FINDINGS 5.2.1 Use of Group I Materials in Applications Important to Structural Integrity of Supports None found.
5.2.2 Thick Section Use of Group II Materials in Applications Important to Structural Integrity None, found.
5.2.3 Thin Section Use of Group II Materials in Applications Important To Structural Integrity ASTM 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 II steel when furnished as formed and without additional specification requirements. However, ASTM A-618 Grade 2 was specified for this tubing and Charpy V-notch testing was required.
Specification ASTM A-618 Grade 2 limits silcon 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 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.
fill FranMln Research Center A Svislon d The Franldln hearne
TER-C5257-167/168 5.2.4 Use of Materials Classified Group III 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 II 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-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.
5.2.5 Use of Materials Classified Group III by NUREG 0577, Outright All bolting and welding materials.
5.2.6 Issues Not Completely Resolved The text and materials table of the IMPC letter of response refer to use of ASTM 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.
This question, although unresolved, would not appear to affect the final if class ication o f this plant.
- 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.
(Ill FranMin Research Center A Division or %he Franl4n Insets t
-0 TER-C5257-167/168 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. In the review, general criteria of NUREG 0577 were specifically applied to information furnished by Indiana and Michigan Power Company (IMPC) 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 sup'port design of other components.
In the case of D. C. Cook Units 1 and 2 FRC concludes that:
1 ~ Engineering measures taken in support design, material selection, 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 III 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 III (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.
I)9 FranMin Research Center f
A Osvlslon er The renk5n Insdrute
0