United States Nuclear Regulatory Commission - Protecting People and the Environment

Generic Technical Activity A-12 Fracture Toughness (Generic Letter 80-46)



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, 

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 1990 

                                   - 2 -

1.   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  material of 
          steam generator skirts; material used in the upper steam generator

     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: 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  in pump supports. 

                                  - 3 -                       May 19, 1980

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

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

     J.   Prairie Island Units 1 and 2: 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 

     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, 

                                  - 4 -                        May 19 1980 

     are to recommend appropriate action and provide a schedule for such 
     action. If any support contains a material not evaluated by  (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: 

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


     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. 

                                  - 5 -                        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. 


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

"General Operating Reactor
  Review Procedure and
  Acceptance Criteria"

cc w/attachment:
See next page

Page Last Reviewed/Updated Friday, May 22, 2015