ML20238F710
| ML20238F710 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/28/1998 |
| From: | Colburn T NRC (Affiliation Not Assigned) |
| To: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| References | |
| GL-97-01, GL-97-1, TAC-M98605, NUDOCS 9809040225 | |
| Download: ML20238F710 (5) | |
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Sh-l 43 UNITED STATES i
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NUCLEAR REGULATORY COMMISSION j
g% * * * * * #j WASHINGTON, D.C. 3066H001 j
August 28, 1998 l
l Mr. James W. Langenbach, Vice President and Director-TMl GPU Nuclear, Inc.
P. O. Box 480 Middletown, PA 17057
SUBJECT:
GENERIC LETTER (GL) 97-01, " DEGRADATION OF CRDM/CEDM NOZZLE AND OTHER VESSEL CLOSURE HEAD PENETRATIONS," REQUEST FOR ADDITIONAL INFORMATION FOR THREE MILE ISLAND NUCLEAR STATION, UNIT NO.1 (TMI - 1) (TAC NO. M98605)
Dear Mr. Langenbach:
On April 1,1997, the staff issued Generic Letter (GL) 97-01, " Degradation of Control Rod Drive Mechanism (CRDM/CEDM) Nozzle and other Vessel Closure Head Penetrations," to the industry requesting in part that addressees provide a description of the plans to inspect the vessel head penetration nozzles (VHPs) at their respective pressurized water reactor (PWR) designed plants.
With respect to the issuance of the GL, the staff required the addressees to submit an initial response within 30 days of issuance informing the staff of the intent to comply with requested information and a follow-up response within 120 days of issuance containing the technical details to the staffs information requests, in the discussion section of the GL, the staff stated that
" individual licensees may wish to determine their inspection activities based on an integrated industry inspection program...," and indicated that it did not object to individual PWR licensees basing their inspection activities on an integrated industry inspection program.
As a result, the Babcock & Wilcox Owners Group (B&WOG) determined that it was appropriate for its members to develop a cooperative integrated inspection program in response to GL 97-01.
The B&WOG program is documented in Topical Report BAW-2?O1, " Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations," which was prepared by Framatome Technologies, incorporated (FTI) on behalf of the B&WOG and the following B&WOG member utilities and plants:
GPU Nuclear, Inc. - Three Mile Island Unit 1 Duke Power Company-Oconee Nuclear Station Units 1,2, and 3 Entergy Operations, Inc. - Arkansas Nuclear One Unit 1 Centerior Energy Corp. - Davis Besse Nuclear Plant Florida Power Corporation - Crystal River Unit 3 The B&WOG submitted its integrated program and Topical Report BAW-2301 to the staff en July 25,1997.
The staff has determined by your letters dated April 30, and July 29,1997, that you were a member of the B&WOG and a participant in the B&WOG integrated program that was developed to address the staffs requests in GL 97-01, in your letters of April 30 and July 29,1997, you also indicated that the information in Topical Report BAW-2301 is applicable with respect to the assessment of VHP nozzles at TMl-1.
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6 J. Langenbach The staff has reviewed your responses to GL 97-01, dated April 30, and July 29,1997, and requires further information to complete its review of the responses as they relate to the B&WOG's integrated program for assessing VHP nozzles at B&WOG member plants, and to the contents of Topical Report No. BAW-2301.
The enclosure to this letter forwards the staff's request for additional information (RA!). We request a response to the RAI within 90 days of receipt of this letter. It should be noted that similar staff reques!J have been issued to the other B&WOG member utilities. As was the staff's position before, the staff encourages you to address these inquiries in integrated fashion with -
the B&WOG; however, the staff also requests that you identify any deviations from the B&WOG's integrated program that may be specific to your facilities. We appreciate your efforts expended with respect to this matter.
Sincerely,
. Original signed by Timothy G. Colbum, Senior Project Manager Project Directorate 1-3 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Enclosure:
Request for AdditionalInformation cc w/ encl: See next page
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DOCUMENT NAME: G:\\COLBURN\\GL9701.RAI To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No co)y 3 in IE, OfD13_.c lM l
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DATE 08/ 1 %' /98 08/ M\\ /98
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'J.Langenbach Three Mile Island Nuclear Station, Unit No.1
- cc:
Michael Ross Robert B. Borsum Director, O&M, TMI B&W Nuclear Tet.bnologies GPU Nuclear, Inc.
Suite 525 P.O. Box 480 1700 Rockville Pike Middletown, PA 17057 Rockville, MD 2085.1 John C. Fomicola William Domsife, Acting Director Director, Planning and Bureau of Radiation Proterction Regulatory Affairs Pennsylvania Department of GPU Nuclear, Inc.
Environmental Resources 100 Interpace Parkway P.O. Box 2063 Parsippany, NJ 07054 Harrisburg, PA 17120 Jack S. Wetmore Dr. Judith Johnsrud Manager, TMI Regulatory Affairs National Energy Committee GPU Nuclear, Inc.
Sierra Club P.O. Box 480 433 Orlando Avenue Middletown, PA 17057 State College, PA 16803 Emest L. Blake, Jr., Esquire Peter W. Eselgroth, Region l Shaw, Pittman, Potts & Trowbridge U.S. Nuclear Regulatory Commission 2300 N Street, NW.
475 Allendale Road Washington, DC 20037 King of Prussia, PA 19406 Chairman Board of County Commissioners of Dauphin County Dauphin County Courthouse Harrisburg, PA 17120 Chairman Board of Supervisors of LondonderryTownship R.D. #1, Geyers Church Road Middletown, PA 17057 Wayne L. Schmidt Senior Resident inspector (TMI-1)
U.S. Nuclear Regulatory Commission P.O. Box 219 Middletown, PA 17057 Regional Administrator Region i U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406
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Request for Additional Information for Utilities Participating in the Babcock and Wilcox Owners Group (B&WOG)
Integrated Response to Generic Letter (GL) 97-01,
" Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations" Applicability of Topical Report Number BAW-2301 to the Plant-Specific Responses to GL 97-01 for Participating Member Utilities and Plants in the B&WOG The methodology developed by Framatome Technology incorporated (FTI) for predicting the susceptibility of vessel head penetration nozzles in B&WOG plant designs is provided in Appendix B to the report, " Description of Control Rod Drive Mechanism (CRDM) Nozzle PWSCC Inspection and Repair trategic Evaluation Model." The CRDM Nozzle PWSCC Inspection and Repair Strategic Evaluation (CIRSE) methodology for crack initiation is dependent on the calculation of a Relative Susceptibility Factor (RSF), which in part is a function of a number of multiplicative adjustment factors (e.g., the material factors, fabrication factors, and water chemistry factors). FTl has assumed that there is little variability in the alloying chemistries and microstructure of the heats used to fabricate the B&W CRDM penetration and thermocouple nozzles, and has therefore set the values for these multiplicative adjustment factors to a value of 1.0. This simplifies the CIRSE crack initiation model to one that is simply based on the applied nozzle stresses and nozzle operating temperatures. The approach taken does not appear to be consistent with the ranges of data provided in Table 1 of the report, "CRDM Nozzle Heats at B&W-Design Plant," which provides the yield strengths, ultimate tensile strengths, and carbon contents for the B&W CRDM penetration nozzle material heats. The data in Table 1 of the report imply that there may be some variability in the chemistries and microstructure of the Alloy 600 material heats used to fabricate the B&W CRDM penetration nozzles.
Topical Report No. BAW-2301 also provides the B&WOG's inspection schedule and scope for VHP nozzles in B&W designed plants. In this section, the B&WOG indicated that the schedule
' for VHP nozzle inspections was developed based on the susceptibility assessments of the B&W CRDM penetration nozzles and thermocouple nozzle heats. The specific results of ine CRDM penetration nozzle susceptibility rankings for the B&WOG plants were not provided in the report;
. however, the B&WOG has indicated that additionalinspections of the B&W fabricated CRDM penetration nozzles have been scheduled for the 1999 refueling outages (RFOs) of the Oconee Nuclear Station Unit 2 (ONS-2) and at Crystal River Unit 3 (CR-3) plants. In addition, FTl has also indicated that additionalinspections of the thermocouple nozzles at Three Mile Island Unit 1 (TMI-1) and Oconee Nuclear Station Unit 1 (ONS-1) are tentatively scheduled for the year 2001.
Therefore, with respect to the design of the CIRSE crack initiation and crack growth models, the susceptibility rankings for vessel head penetrations in B&W designed plants, the proposed CRDM nozzle inspections at ONS-2 and CR-3, and the postulated inspections of the instrumentation nozzles at TMI-1 and ONS-1, the staff requests the following information:
- 1. Provide a description of how the various product forms, material specifications, and heat treatments used to fabricate each CRDM penetration nozzle at the B&WOG member utilities are handled in the C,lRSE model.
ENCLOSURE
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- 2. Provide any additional information, if available, regarding how the model will be refined to allow the input of plant-specific inspection data into the model's analysis methodology.
- 3. Describe how FTl's crack initiation anb crack growth models for assessing postulated flaws in vessel head penetration nonles were bench-marked, and a listing and discussion of the standards the models were bench-marked against.
- 4. Provide the latest CIRSE model susceptibility rankings of B&W designed facilities based on the CIRSE model analysis results compiled from the analyses of the CRDM and instrumentation nonles at the facilities.
- 5. Compare the overall susceptibility rankings of the thermocouple nonles at TMI-1 to that of the plants with the most susceptible ranked CRDM penetration nonlos. Based on this assessment, indicate whether the thermocouple nozzles at TMI-1 will be inspected during the year 2001 refueling outage, if it is determined that the thermocouple nonles will not be inspected, provide the basis for omitting the inspections of the thermocouple nonles in the year 2001.
- 6. Given that the TMI-1 facility experienced an extended intrusion of thiosulfate ions into the TMI-1 RCS, and since the degradation of Alloy 600 steam generator tubes at TMI-1 has in part been attributed to this event, justify why the Alloy 600 CRDM penetration noules at TMI-1 are not being scheduled for volumetric inspection in the near term.