ML053070380
| ML053070380 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/08/1998 |
| From: | Hansen A Office of Nuclear Reactor Regulation |
| To: | Jeffery Wood Centerior Service Co, Toledo Edison Co |
| References | |
| GL-97-001, TAC M98561 | |
| Download: ML053070380 (3) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINOTON, D.C. -1 September 8, 1998 Mr. John K. Wood Vice President - Nucl ear, Davi s-Besse Centerior Service Company c/o Toledo Edison Company Davis-Besse Nuclear Power Station 5501 North State Route 2 Oak Harbor, OH 43449-9760
SUBJECT:
GENERIC LETTER 97-01, "DEGRADATION OF CRDM/CEDM NOZZLE AND OTHER VESSEL CLOSURE HEAD PENETRATIONS,"
REQUEST FOR ADDITIONAL (TAC NO. M98561)
INFORMATION, DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1
Dear Mr. Wood:
On A p r i l 1, 1997, the s t a f f issued Generic Letter (GL) 97-01, "Degradation o f CRDM/CEDM Nozzle and Other Vessel C1 osure Head Penetrations," t o the industry requesting i n part t h a t addressees provide a description o f the plans t o inspect the vessel head penetration nozzles (VHPs) a t t h e i r respective pressurized water reactor (PWR) plants.
GL, the s t a f f required the addressees t o submit an i n i t i a l response within 30 days o f issuance informing the staff o f the i n t e n t t o provide requested information and a follow-up response within 120 days o f issuance containing the technical d e t a i l s t o the s t a f f ' s information requests.
I n the discussion section o f the GL, the s t a f f stated t h a t "individual licensees may wish t o determine t h e i r inspection a c t i v i t i e s based on an integrated industry inspection program...,I' and indicated t h a t it d i d not object t o individual PWR licensees basing t h e i r inspection a c t i v i t i e s on an integrated industry inspection program.
With respect t o the issuance o f the 0
As a result, the B&W Owners' Group (B&WOG) determined t h a t it was appropriate f o r i t s members t o develop a cooperative integrated inspection program i n response t o GL 97-01.
The B&WOG program i s documented i n Topical Report BAW-2301, "Degradation o f CRDM/CEDM Nozzl e and Other Vessel C1 osure Head Penetrations, 'I which was prepared by Framatome Techno1 ogi es, Incorporated (FTI) on behalf o f the B&WOG and the following B&WOG member u t i l i t i e s and p l ants :
General Public U t i l i t i e s - 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 Power Station, Unit 1 Florida Power Corporation - Crystal River, Unit 3.
The B&WOG submitted i t s integrated program and Topical Report BAW-2301 t o the s t a f f on July 25, 1997.
Based on your responses dated April 23 and July 28, 1998, concerning the a
Davis-Besse Nuclear Power Station, the staff has determined that you are a participant in the B&WOG integrated program that was developed to address the staff's requests in GL 97-01. In your responses, you also indicated that the information in Topical Report BAW-2301 is applicable with respect to the assessment of VHP nozzles at Davis-Besse.
The staff has initiated a review of your responses and requires additional information to complete the review.
The methodology devel oped by Framatome Techno1 ogy Incorporated (FTI) for predicting the degradation susceptibility of vessel head penetration nozzles in B&WOG plant designs is provided in Appendix B to the report, "Description of CRDM Nozzle PWSCC Inspection and Repair Strategic 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 (for example, the material factors, fabrication factors, and water chemistry factors).
FTI has assumed that there is little variability in the alloying chemistries and microstructures 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 microstructures 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.
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 o f the CRDM penetration nozzle susceptibility rankings for the B&WOG plants were not provided in the report. However, the B&WOG has indicated that additional inspections of the B&W-fabricated CRDM penetration nozzles have been scheduled for the 1999 refueling outages (RFOs) of Oconee Nuclear Station, Unit 2 (ONS-2) and Crystal River, Unit 3 (CR-3) plants. In addition, FTI has also indicated that additional inspections of the thermocouple nozzles at Three Mile Island, Unit 1 (TMI-1) and Oconee Nuclear Station Unit 1 (ONS-I) are tentatively scheduled for the year 2001.
0 In this section, the B&WOG
Therefore, with respect to the design of the CIRSE crack initiation and crack e
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 are handled in the CIRSE model.
- 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 models analysis methodology.
Describe how FTIs crack initiation and crack growth models for assessing postul ated fl aws in vessel head penetration nozzles were bench-marked, and provide a listing and discussion of the standards the model s were bench-marked against.
- 3.
- 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 nozzles at the facilities.
Please respond to this request within 90 days of the date of this letter.
Similar staff requests are being issued to the other B&WOG member utilities.
The staff encourages you to address these inquiries in integrated fashion with the B&WOG.
B&WOGs integrated program that may be specific to your facility.
The staff also requests that you identify any deviations from the Si ncerel y,
Allen G. Hansen, Project Manager Project Directorate I 11-3 Division of Reactor Projects III/IV Office of Nucl ear Reactor Regul ati on Docket No. 50-346 cc: See next page