ML19337A846

Fracture Toughness of Steam Generator & Reactor Coolant Pump Supports,Inidian Point Unit 2, Interim Technical Evaluation Rept

ML19337A846
Person / Time
Site:	Indian Point
Issue date:	09/26/1980
From:	Allten A, Noell P, Stilwell T FRANKLIN INSTITUTE
To:	Fair J Office of Nuclear Reactor Regulation
Shared Package
ML100340720	List: ML100340719 ML19337A846
References
CON-NRC-03-79-118, CON-NRC-3-79-118 TER-C5257-169, NUDOCS 8009300433
Download: ML19337A846 (17)

Text

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! O (INTERIM) i TECHNICAL EVALUATION REFORT FRACTURETOUGHNESS OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS

CONSOLIDATED EDISON COMPANY OF NEW YORK, INC.

INDIAN POINT UNIT 2 NRC DOCMETNO. 50-247 NRC TAC NO. 08776 FRC PROJECT C5257 NRC CONTRACT NC. NRC 03 75118 FRC TASK 169 Preparedby 4 Franklin Research Center Authors: T.c.Stilwell, P.N.Noell, The Parkway at Twentieth Street A.G.Alleen, K.E.Dorschu Philadelphia, PA 19103 FRCGroup Leader: T.c.Stilwell Preparedfor NuclearRegulatory Commission Washington, D.C. 20655 Lead NRC Engineer: J.R. Fair September, 1980 This report was prepared as an account of work sponsored by an agency of me United States Government. Neimer the United States Govemment nor any agency moreof, or any of meir employees, makes any warranty, expressed or implied, or assumes any legal liability or responsleility for any t?:rd party's use, or tne results of such use, r any information, apparatus, product or process disclosed in 41s report, or represents that its use by such third party would id Infringe prtretely owned rfghts.

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CONTENTS Section Title Page 1

SUMMARY

. . . . . . . . . . . 1 2 INTRODUCTION. . . . . . . . . . 2 3 BACKGROUND . . . . . . . . . . 3 4 CRITERIA APPLIED IN THE EVALUATION . . . . . 5 4.1 Fracture-Toughness Grouping of Materials Used in Support Construction . . . . . . . 5 4.2 Plant Grouping for Fracture-Toughness Ranking of S/G and RCP Support Structures . . . . . 6 4.3 Criteria for Fracture-Toughness Adequacy of S/G and RCP Supports . . . . . . . . 6 5 TECHNICAL EVALUATION . . . . . . . . 7 5.1 Use of Group I Materials . . . . . . 10 5.2 Use of Group II Materials in Thick sections . . 10 5.3 Use of Other Metals of Problematic Fracture-Toughness . . . . . . . . . 11 5.4 Insufficient Information on Welding Practices . . 11 5.5 Comments on the Stress Summary . . . . . 12 6 CONCLUSIONS . . . . . . . . . 13 TABLE Number Title Page 5.1 COMPONENT SUPPORT

SUMMARY

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FRACTURE TOUGHNESS OF STEAM GENERATOR AND REACTOR COOLANT PUMP SUPPORTS

1.

SUMMARY

Information concerning aspects of the fracture-toughness design of the steam generator (S/G) and reactor coolant pump (RCP) supports for the Indian Point Unic 2 nuclear power station was submitted to the Director of Nuclear Regulation by the Consolidated Edison Company of N.Y., Inc. (CEC) by letter dated May 31, 1979. This information was reviewed at the Franklin Research Center (FRC) and evaluated in accordance with the criteria of the Nuclear Reg-ulatory Connaission (NRC) as set forth in NUREG 0577-Draft (henceforth referred to simply as NUREG 0577).

FRC found that the design of the supports for the RCP and S/G incorporates memoers ordered to material specifications classified by NUREG 0577 as Group I (i.e. , specifications governing products of relatively poor fracture tough-ness). In particular, seamless pipe ordered to the American Society for Test-ing and Materials (ASTM) specification A-53 is used for major members of the RCP support frame and in the S/G support structure also.

In addition, FRC found that these supports have major structural members witn thick cross sections ordered to ASTM A-36 and ASTM A-108 Grade 1018.

NURIC 0577 assigns a Group II (intermediate fracture toughness ranking) to ASTM A-36 and also to steels comparable to ASTM A-108 Grade 1013. Although ASTM A-108 Grade 1018 is not ranked by NUREG 0577, it is FRC's judgetat that it clearly merits a Group II ranking.

NUREG 0577 provides nil ductility temperature (NDT) screening criteria by which the fracture toughness suitability of materials for use in S/G and RCP l support structures can be quickly appraised. According to these criteria, use l 42 ..d Franklin Research Center 4 om e n. re

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TER-C5257- 169 (Interim) of these Group II materials in thin structural members is acceptable, but j general use in thick members is not sanctioned. Specifically, the questioned applications are:

1. Thick flanged I-beams of ASTM A-36 steel which are used for the main vertical and cross support members of the S/G support'.

2. Tie rods of 4 /4-in 3 diameter, made from ASTM A-108 Grade 1018 steel, which provide horizontal restrain: to 4

the RCPs.

Additional information relating to welding design, procedures, and in-spections is required before a definitive judgment of their fracture-toughness adequacy can be rendered.

It may be possible to demonstrate the fracture-toughness adequacy of the questioned components by other methods. Acceptable alternative methods for such evaluations are discussed in NUREG 0577.

Pending such demonstrations, FRC recommends that a tentative Group I plant classification for fracture toughness of S/G and RCP supports be, assign-ed to Indian Point Unit 2.

2. INTRODUCTION This report provides a technical evaluation of information supplied by CEC with its letter of May 31, 1979, to the Director of Nuclear Reactor Regu-lation. The information concerns the fracture-coughness design of supports for the S/Gs and RCPs for Indian Point Unic 2. The objective of the evalua-tion is to rank the design for fracture-toughness integrity on a relative scale in accordance with the grouping scheme and criteria established in NUREG l 0577.

I The ranking is considered tentative because: l

1. It is based solely on review of the information submitted.

2. NUREG 0577 and the criteria it contains was not developed when the j information was requested from the licensee. Additional, more I specific information relevant to the plant grouping may be on hand and might have been submitted had NUREG 0577 been available to provide guidance at the time that the information was solicited.

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3. BACKGROUND During the course of the NRC licensing review for two pressarized water reactors (PWR), North Anna Units 1 and 2, questions vers raised regarding the fracture-tmaghness adequacy of certain members of the S/G and RCP supports.

The potential for lamellar tearing in some mapport members was also questioned.

The staff's concern in the North Anna licensing process was that not enough attention might have been paid 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. Tou ghnes s also implies that under specified conditions the matarial has the capability to arrest the grcwth 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 rem 21t in failure of the mapport structure under postulated accident conditions, specifically, loss-of-coolant accident (LOCA) and safe shutdown earthquake (SSE).

To address fracture teughaess concerns at the North Anna facility, the licensee undertook tests not originally specified and not included in ebe relevant ASTM specifications. These tests indicated that material used in certain mapport members had relatively poor fracture toughness at 80*F metal tempera tu re.

In this case the licensee agreed to raise (by ancillary electrical heat) the temperature of the S/G support beams in question to a minimam of 225 ?

avery time, throughout the life of the plant, the reactor coolant system (RCS) is presmarized above 1000 psig. The NRC staff fotnd this to be an acceptable resolu tion.

Because similar materials and designs , e used in other plants and be-cause similae problems were therefore posr. this matter was incorporated into the NRC Program for Resolution of Generic Issues as " Generic Technical

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Activity A-12-- Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pu'ap Supports."

Since the original licensing action (North Anna Units 1 and 2) nvolved only the S/G and RCP mapports of PWRs, the staff's initial ef forts were di-rected toward examination of the corresponding m2pports at other PWR facili-ties. However, the staf f has kept in mind the possibility of expanding its review to include other support senaccures in PWR plants and support struc-cures in boiling water reactor (BWR) plants.

The integrity of support embedments was not questioned during the North Anna licensing action, and emphasis was consequently placed on resolving the most immediate generic ismae--whether or not problems similar to those uncov-cred at Nor,th 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 coasiderations were deemed more m2lted 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 62 ring l978.

Sandia Laboratories of Albuquerque, New Mexico, was retained to assist the staf f in the review and analysis of the information received from licens-ees and applicants. Based on an analysis of the information, the technical sb2 dies made by Sandia Laboratories, and review of the issues by the NRC staff, the NRC developed an NRC staff technical position on these issues.

This 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 m2mber of PWR plants to achieve plant grespings according to the relative fracture-tmaghness inte-grity of the,e m2pports. The plan,t ratings are:

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. e Group I (lowest) e Group II (intermediate) e Group III (highest)

NUREG 0577 also emphasizes the tentative character of these rankings, acunowledging that a nuncer of plants were classified as Group I becasue li-consees had not submitted all the information needed for definitive classifi-cation. Therefore, they had not demonstrated that their plant merited a high-er ranking. In this regard NUREG 0577 states:

Receipt of such (i.e., currently unsubmitted) information could result in the plants being moved to a lower suscept-ibility (to brittle failure) group af ter very little addi-tional analysis.

The reply to the NRC request was received too late from Indian Point for detailed review prior to issuance of NUREG 0577. However, based on informa-tion then available to NRC, a preliminary assessment was made leading to as-signment of a plant group ranking in NUREG 0577.

The present evaluation applies the criteria of NUREG 0577 to the informa-tion received concerning the fracture-toughness design of the S/G and RCP sup-ports to provide an independent assessment based upon the evidence submitted.

4. CRITERIA APPLIED IN THE EVALUATION 4.1 FRACTURE-TOUGHNESS GROUPING OF MATERIALS USED IN SUPPORT CONSTRUCTION 4.1.1 Criterion Table 4.6--Materials Groups--of Appendix C to NUREG 0577 groups materials according to their relative fracture toughness as:

e Group I (poorest) e Group II (intermediate) e Group III (best) 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 superior

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TER-C5257-169 (Interim) to those routinely supplied under the ASTM (or other standard) specification, then the material was grouped in accordance with Table 4.6.

If additional requirements aimed at procuring a product with superior fracture-tcushness properties were specified, consideration was given to cred-iting this specific material order with an improved material-group rating.

4.2 Pl. ANT 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 w.ithin 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, provided this usage is important to support integrity.

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 cake appropriate corrective measures to assure their fracture-toughness integrity.

NUREG 0577 provides guidance for such demonstrations.

4.3.1 NDT Criteria for Screening i30*F NDT + 1. 3 8 + 'or g Tsupports('F)

,60*F where:

e 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.

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a is the standard deviation for the data used to determine NUY as listed in Table 4.4.

e port is the lowest metal temperature that the support Tsu$er mem w,ill ever experience throughout the plant life when the plant is in an operational state. In the absence of measured, plant-specific data, Tsupports is taken as 75'F.

e 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, 3, or T, pp, g, are actually valid for the S/G or RCP supports and materi-als in the licensee's plant, such data may be used. If acceptable alternative evidence is not available, the values stipulated above should be used.

4.3.3 Alternative Criteria NUEEG 0577 also recognizes that fracture-toughness integrity is a complex matter involving a number of interrelated factors, most of which are plant-s pecific. 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 showing compliance with the intent of NUREG 0577 are possible, but also offers extensive guidance relating 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, in-stead, encourages each licensee to review the fracture-toughness adequacy of his S/G and RCP supports and submit evidence of his findings.

5. TECHNICAL EVALUATION A review has been made of information contained in the response dated May 31, 1979, from Mr. W. J. Cahill, Jr. to a request for information, dated nk!!n Research Center A DMeson of The Frannan insatute l

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September 14, 1977, from Mr. A. Schwencer of the NRC. The request sought in-formation concerning the fractt:re toughness of, and the potential for lamellar tearing in, the S/G and RCP supports. A copy of this request (in generic form) may be found in NUREG 0577, Appendix B. Key items from CEC's response were condensed to tabular form and are presented as Table 5 1.

FRC's review addressed only the fracture-toughness issues. Items found to be of concern are briefly summarized below and then discussed more fully in the following eactions of the report.

Use of Group I Materials. ASTM A-53 seamless pipe classified by NUREG 0577 as a Group I (relatively poor fracture toughness) material is used for major members of the RCP support frame. The stresses reported in some of these members during SSE and LOCA are substantial.

Use of Group II Materials. I-beams with thick flanges formed from A-36 steel are used as major members of the S/G supports. ASTM A-36 steel is clas-sified by NUREG 0577 as a Group II material. The NDT screening criteria pro-vided by NUREG 0577 show thin-section applications of A-36 steels to be ac-ceptable for use in S/G and RCP supports but do not sanction use of A-36 in thick sections.

Use of Materials of Problematic Fracture Toughness. The tie rods providing horizontal restraint for the RCPs are 4 3/4-in in diameter and are formed from ASTM A-108 Grade 1018 steel. FRC reconunends a Group II fracture toughness ranking for this steel. ASTM A-108 Grade 1018 meets the NUREG 0577 screening criteria when used in thin members but fails it when used in thick members. The tie rods are unquestionably both thick and important to struc-tural integrity.

Insufficient Information on Welding Practices. The response to items in the NRC's generic letter relating to welding design, welding procedures, nklin Research Center A OMe an of The Frannhn inseeuse

TAM E 5.8 CLAIPODElff SUPPORT SumARY PLANT: INDIAN POINT Uhlf 2 E

trffLITY NSSS AE SUPPOkT SUPPLIER D ~tt conselsdated EJieon United Engineers aaJ Pittsburgh Bridge & tron (PRI)

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O3 f3 MATLEIAl.S MAIIMuse AltJJWattE DESIGN STkESS T FRACTURE MILL CERTS. NEAT NDE 048 TouCloseESS MEMBRAME & TIISDUGli g TYPE AVAILABLE TREATMENT MATERIAL TEST kEIIDIhG (IIORMAL) THICK 41ESS

{ A-36 Yes, (not - A-Sl4, crede F -

ham. value for all loads A-514, Csade F turaished) reactor coolant pump A-53, Csede 5 (D + 3 + A) = 30.0 he t Bolte, Pisa. BoJs A-325 or A-490 Alst 4340 A-108. Crede 1018 AISI 4340 W lJ anatoriales quenched &

SMAW: E7010. E7028; tempered FCAW: Oneswtgon 163 Flus-Cored Wire.

SAW: Lincola 1.60 Wire, Lincoln 780 rius FABBICATION teETHODS USED TO WEI.blNG WELDING FOST-WELDillC PREVENT LAMELLAR IIDE AleD IleSPECTl0Ils PaocESS PROCEDuas TREATMENT TEAhlleG PEkFORMED aihielded Metal AWS Dl.0-66 SC A-Sl4 Cr. F Cood Widing All shop / site welds, Arc (SMAW) Welds, $5:3 Practices" Magnetic Particle or Flus-Cored Arc AWS Sti-65 Suprosts, are Liquid Penetreet Pro-(FCAW) heat treated cedure (LP-l) RCP sup-D Submerged Arc (SAW) AWS DI.0-66 port #21 8001 visually inspected 4/76.

DEsicu TYPE OF SUPPOkT CODE IfSED l.DADEIGC C000D571000S MIlllteust TEMPERATURE OF SUPPOST Specu leams (fus -

For SC 1. 80erenal (D + P + T) 70*F Slautd a n; 90* to 120*F SC and itCP & BCP 2. Blormal + Seismic Operation (but never directly supporte) Supporte 3. Nosmal + Man Seismic measured)

4. Ilormal + Pipe Rupture (acclJeut )

Es5D TER-C5257-169 (Interim) and weld inspections was .not developed to a depth sufficient to permit assess-ment of weld fracture-toughness adequacy.

5.1 USE OF GROUP I MATERIALS Seamless pipe ordered to ASTM specification A-53 Grade B is used in the hot finished condition for major members of the RCP frame structure and also in the S/G supports. This specification provides lictie control over the fracture-toughness characteristics of the product, and consequently, the mate-rial is classified as a Group I (relatively poor fracture toughness) steel by NUREG 0577. ASTM A-53 Grade B also fails to meet the NUREG 0577 NDT screening criteria for use in RCP and S/G support structures.

Stress data for individual frame members were furnished for a aembination of deadweight, seismic, and accident loads. Membrane stress levels in several RCP members constructed of A-53 steel are shown as near yield stress and mem-brane plus bending stresses (elastically computed) are shown to exceed yield strength substantially in some A-53 members. Thus, in this design it appears that a material of questionable fracture-toughness characteristics, A-53 Grade B steel, is used in structural members where computed design stresses are high. The need for a careful and detailed fracture-toughness evaluation is evident.

5.2 USE OF GROUP II MATERIALS IN THICK SECTIONS Structural shapes of ASTM A-36 steel are used extensively throughout the S/G and RCP support structures.

ASTM A-36 is classified by NUREG 0577 as a Group II material (i.e. , one of intermediate fracture toughness). According to the NUREG 0577 NDT screen-ing criteria, use of A-36 in thin lections is acceptable; but its use in sup-port members with thick sections n.y not be acceptable.

Thick sectiona are found in the flanges of a number of I-beams used as principal members of both the S/G and RCP supports. Examples of these struc-tural members are I-beams of the following sizes:

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Since support members which are thick flanged I-beams formed from A-36 steel do not meet the requirements of the screening criteria given in NUREG 0577, the fracture-toughness adequacy of such members must be established by other means. NUREG 0577 provides guidance with respect to acceptable methods for making such evaluations. Documentation of the fracture-toughness adequacy of the thick A-36 members by any one of these methods is necessary before this usage can be dismissed from concern.

5.3 USE OF OTHER METALS OF PROBLEMATIC FRACTURE TOUGHNESS Horizontal restraint of the reactor coolant pumps is provided by tie rods m.ade of ASTM A-108 Grade 1018 steel.

Although this material specification is not explicitly assigned a fracture-toughness ranking by NUREG 0577, companion steels of similar charac-teristics (e.g. , AISI 1017 and AISI 1020) are rated Group II (intermediate fracture toughness). In FRC's judgment, AS.TM A-108 Grade 1018 likewise merits a Group II ranking.

As a Group II steel, ASTM A-108 Grade 1018 can be expected to meet the NUREG 0577 NDT screening criteria if used in thin sections, and fail these criteria if used in thick sections.

The tie rods are 4 3/4-in in diameter, which is clearly thick-section i use. Consequently, the 4 3/4-in, ASTM A-108 Grade 1018 tie rods must be re-gr.rded as candidates for possible brittle behavior, particularly if they were tc experience the dynamic loadings from a LOCA. Therefore, definitive assess-ment of the fracture-toughness adequacy of these tie rods is needed.

l 5.4 IN"UFFICIENT INFORMATION ON WELDING PRACTICES The response to Item 6 of the NRC's generic inquiry states that "the three welding processes (shielded metal arc, submerged arc, and flux-cored arc) were used as required to balance shop flow of the various carbon steel i

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TER-CS257-169 (Interim) weldments." Although this names the welding process used, it fails to provide information addressing precautions taken to minimize residual stress and as-sure fracture toughness of weldsents. For example, was submerged are welding used for economic reasons in some cases where a balanced lower welding-energy input process would have produced lower residual stresses? In general, there is insufficient information provided to demonstrate that reasonable precautions were taken to minimize residual stresses, reduce the possibility of cracking, and assure fracture toughness.

5.5 COMMENTS ON THE STRESS

SUMMARY

The submittal states that normal, upset, emergency, and faulted loading terminology is not applicable to Indian Point Uni c 2, (i.e., the design of the supports predates issuance of Regulatory Guide 1.48) and, therefore, need not necessarily conform to these requirements which are currently mandatory for recently constructed plants. Nevertheless, stresses were provided for simul-taneous loading from dead weight plus seismic load plus pipe rupture. Thus, the stress data furnished appear to be in harmony with the intent of present design requiremants. Moreover, the licensee points out that the stress data as furnished may, in some instances, ue conservative (i.e., reported stresses exceed actual stresses). This conservatism comes about because the stresses were computed as the sum of the most severe stresses produced by each contri-buting load when it alone acts on the structure. Since maximum stress effects from all loadings are not necessarily simultaneous, conservatism may result from the way that the stresses were added. Under conditions of normal opera-tion, reported stresses for principal members of both the S/G and RCP supports are low (ranging from a maximum of about 1/3 of yield strength downward).

Consequently, under ordinary circumstances (normal operation) there appears to be little need for concern about fracture-toughness adequacy.

However, the fracture-toughness issue centers on the adequacy of the sup-ports to sustain a possible (once-in-a-reactor-lifetime) load which might occur under faulted loading conditions. The reported stresses for conditions corresponding to faulted loading span a much broader range. Membrane stresses of about one-half yield strength occur in several members of the S/G supports, l l

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TER-C5257-169 (Interim) and stresses near yield are shown for members of the RCP supports. Moreover, elastically computed meanrane plus-bending stresses in excess of yield are found in a number of members in both support structures.

All stresses, as reported, appear to meet applicable requirements of Section III of the ASME Boiler and Pressure Vessel Code. However, this fact alone is not adequate evidence of fracture-toughness integrity. On the con-trary, the reported stress results show the presence of substantial membrane stresses and membrane plus-bending stresses which (computed on an elastic basis) exceed yield in materials, specifically ASTM A-53 and A-36, which do not ordinarily exhibit superior fracture-toughness characteristics. This clearly points to the need for more detailed investigation of such members with respect to their fracture-toughness adequacy. .

6. CONCLUSIONS Information received from CEC relating to the fracture-coughness design of the S/G and RCP pump supports for Indian Point Unit 2 has been reviewed at FRC and evaluated in accordance with the criteria of NUREG 0577.

It was found that the supports include members made of steels ordered to specifications classified by NUREG 0577 as Group I, the fracture toughness of which is relatively poor. In Indian Point Unit 2, ASTM A-53 seamless pipe (a Group I material) is used for major members of the RCP support frame and in the S/G structure also.

In addition. steels ranked (either explicitly or implicitly) by NUREG 0577 as possessing intermediate fracture toughness (Group II materials) were used in members having thick sections. Such applications occur in the thick flange I-beams of ASTM A-36 steel used for the main verticals and cross sup-port members of the S/G support frame and the 43/4-in diameter tie rods of ASTM A-108 Grade 1018 steel used to provide horizontal restraint for the RCPs. When the NDT screening criteria of NUREG 0577 are applied to these steels, it is found that the criteria permit their use in thin S/G and RCP support members but not (without further qualification) in thick members.

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The information supplied concerning veld designs, welding practices, and .

weld inspections was also revieni but found to be insufficient to permit fracture-toughness evaluation.

Based upon 1) the use of Group I materials, 2) use of Group II materials in thick sections, and 3) provision of information insufficient to demonstrate the fracture-coughness adequacy of welding, FRC reccessends that a tentative Group I plant ranking for fracture-toughness of S/G and RCP supports be as-signed to CEC's Indian Point Unit 2 nuclear power station.

Further in-depth evaluation of these concerns should be undertaken. It is possible that the fracture-toughness adequacy of some or all of the areas in question can be demonstrated by methods described in NUREG 0577 as accept-able to the NRC.

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