NRC Generic Letter 80-46, Generic Technical Activity A-12 Fracture Toughness

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GL80046

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May 19 1980 Docket ACRS

(16) NRC PDR CR BCs LOCAL PDR L;ADs NRR Reading L:LAs TIppolito DEisenhut SNorris RPurple ALL POWER REACTOR LICENSEES Jolshinski RSnaider Atty, OELD

(5) VNoonan OI&E

(5) KWichman NSIC DSellers TERA TNovak

In November 1979, you were sent, for your review and comment, the "For Comment" edition of NUREG-0577, "Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports".

The report provides the staff's resolution of the NRC's Generic Technical Activity A-12, which is an "Unresolved Safety Issue" pursuant to Section 210 of the Energy Reorganization Act of 1974.

The generic study resulted from questions raised during the licensing review of two pressurized water reactors (PWRs). The specific concern was the capability of the supports to maintain their structural integrity under severe environmental and accident conditions.

NUREG-0577 describes the technical issues, the technical studies performed by an NRC consultant, the NRC staff's tentative plans for implementing its technical positions. It also provides recommendations for further generic research into the subject of lamellar tearing. The Electric Power Research Institute has been requested to conduct such research.

Comments on the report, including the proposed review procedure and implementation schedule contained in this letter, are being solicited from interested organizations, groups and individuals. These comments will be evaluated by the NRC staff prior to final implementation of this subject. all comments should be forwarded to Mr. Richard Snaider, Generic Issues Branch, Nuclear Regulatory Commission, Washington, D.C. 20555, by July 7, 1980.

At the completion of the 60 day period, the staff will evaluate the components received and, if needed, will issue a supplement or revision to NUREG-0577. The target for issuing the supplement or revision is late September 1980.

Subsequent to issuance of NUREG-0577, the NRC has decided to modify significantly the implementation plan presented in Section 4 of NUREG-0577. Basically, we propose that licensees be tasked with demonstrating the adequacy of the support structures of their facility(ies) from a fracture toughness standpoint. They will then be required to submit a report to the NRC describing in detail the conclusions drawn and any action taken or planned. The following steps will be required under the proposed implementation plan:

.

May 19 19901. Licensees of operating PWRs should refer to Section 3 (pages 9 and 10)

of NUREG-0577 to ascertain their classification with regard to potential susceptibility to low fracture toughness. Plants not reviewed during the generic study are already undergoing a complete review by the Franklin Research Center (FRC). This includes Arkansas Nuclear One Unit 2. San Onofre Unit 1, and Turkey Point Units 3 and 4. Indian Point Units 2 and 3, which are Group I plants, are included in the FRC review. The NRC technical contact for the FRC review is C. D. Sellers.

2. These licensees in Group III are not required to take any further action with regard to steam generator and reactor coolant pump supports and may consider this matter resolved. Evaluation of other supports should be performed as set forth in Step 6.

3. These licensees in Group II are not required to take any action at this time with regard to steam generator and reactor coolant pump supports. Future action on these supports may be necessary based on information received from the reports of Group I licensees. The NRC staff will make this determination and inform Group II licensees if any additional action is necessary for those classes of equipment. However, these licensees should implement Step 6 for action concerning other supports.

Group I licensees, with the exception of Indian Points Units 2 and 3, are to undertake plant specific evaluation. The specific questions of Appendix D to NUREG-0577 are to be disregarded. The materials/parts listed below are to be analyzed in accordance with the General Operating Reactor Review Procedure enclosed as Attachment 1.

A. Crystal River Unit 3: A-515 in the flange and <ur> material of steam generator skirts; material used in the upper steam generator supports

B. Davis-Besse Unit 1: A-36; A-516, A-515 and A-53 in steam generator lower lateral supports.

C. J. M. Farley Units 1 and 2: Carpenter Custom 455 steel bolts used in Clevis attachments of the vertical columns.

D. Fort Calhoun: A-307 nuts and bolts.

E. Kewanee: <ur>250 CYM 0.5-inch diameter "Heli-coil screws into S. G." and 1.0-inch diameter "upper support ring girder wall bolts"

F. Maine Yankee: Steam generator base castings.

G. Millstone Unit 2: A-106 and A-515 steel <ur> in pump supports.

.May 19, 1980

H. Palisades: A-212 in the base flange of the steam generator supports.

I. Point Beach Units 1 and 2: 12-inch diameter A-53 Schedule 100 pip columns.

J. Prairie Island Units 1 and 2: <ur>250 CYM 0.5-inch diameter "heli-

coil screws into S. G." and 1.0-inch diameter "upper support ring girder wall bolts".

K. Rancho Seco: Materials used in coolant pump horizontal supports; A-515 base flange on steam generator support skirts; materials used in steam generator upper horizontal restraints.

L. St. Lucie Unit No. 1: A-515 in coolant pump snubber clevises; A-

27 in steam generator base castings.

M. Surry Units 1 and 2: Vascomax 300 and 350 steels in all location; A-106 pipes; A-235 plants; A-105 pipe and forgings.

N. Three Mile Island Unit 1: Materials of coolant pump horizontal supports A-515 base flange on steam generator horizontal restraints.

O. Yankee Rowe: Materials used in upper part of steam generator support structures; material used in coolant pump hanger rod supports; A-7 and C-1020 steels.

If the support materials cannot be shown to have adequate fracture toughness or the capability to withstand stress corrosion cracking, licensees must immediately inform the applicable regional office of the NRC's Office of Inspection and Enforcement (OIE), with copies to Director, OIE and Director, Office of Nuclear Reactor Regulation. In this report, licensees are to recommend appropriate action and provide a schedule for such action. All Group I licensees, including Indian Point Units 2 and 3, must also perform the evaluations of additional supports as set forth in Step 6 below.

5. PWR licensees whose plants were not included in NUREG-0577 must also perform the review of steam generator and reactor coolant pump supports. These licensees should begin with the materials classification of Table 4.6 of Appendix C to NUREG-0577 (page C-38) and proceed with the evaluation call for in Attachment 1. If adequate fracture toughness or the ability to withstand stress corrosion cracking (where applicable) cannot be demonstrated, the licensee must immediately inform the applicable regional office of NRC's Office of Inspection and Enforcement (OIE), with copies to Director, OIE, and Director, Office of Nuclear Reactor Regulation. In this report, licensees

.May 19 1980

are to recommend appropriate action and provide a schedule for such action. If any support contains a material not evaluated by <ur> (Table 4.5 of Appendix C), the licensee must perform an evaluation of this material and demonstrate the adequacy of its toughness in the particular application. These licensees must also perform the evaluation of additional supports as set forth in Step 6.

6. Licensees of boiling water reactors (BWRs) PWRs must review the material of applicable major component supports not included in the NUREG-0577 review and determine if the analysis of Attachment 1 is called for. For example, reactor vessel support members of material listed in Group I of Table 4.5 (page C-38) of NUREG-0577 certainly should be reviewed. These supports are:

BWR

Reactor vessel Reactor coolant recirculation pumps (non-pipe supported)

PWR

Pressurizer Reactor vessel

The review method of Step 5 above is to be utilized for those materials deemed by the licensee to require analysis. The report at the completion of this review (see 7. below) must present the results of such analysis or the rationale behind exemption of support materials from analysis.

7. The date for completion of the evaluations, and the submittal of detailed reports discussing methods used, results of evaluations, and subsequent actions to resolve problems, is December 31, 1981. As with the reports discussed above, submittal should be made to the Director of the Regional Office, OIE, with copies to Director, OIE, and Director, Office of Nuclear Reactor Regulation. If any support material was evaluated to have inadequate fracture toughness, the licensee must state what appropriate action has been taken or provide a schedule for such action. The NRC staff will use the information of these reports to determine if any additional action is necessary, in particular on Group II plants. The information will also be used to determine if inservice inspection of supports is necessary.

.May 19 1980

Because of the importance of resolving this issue, plants undergoing Systematic Evaluation Program (SEP) review must be reviewed and any necessary changes implemented in the same time frame as non-SEP plants.

Sincerely,

Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation

Attachment:

"General Operating Reactor Review Procedure and Acceptance Criteria"

cc w/attachment:

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