ML20199E161
| ML20199E161 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 01/15/1999 |
| From: | Krich R COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-01, GL-97-1, NUDOCS 9901200340 | |
| Download: ML20199E161 (7) | |
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Commonwealth Edison Company i
- 1400 Opus Place Downers Grove. IL 60515 5701 January 15,1999.
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk j
Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 i
Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-45Z f-Byron Station, Units 1 and 2 f
Facility Operating License Nos. NPF-37 and NPF-86 NRC Docket Nos. STN 50-454 and STN 50-455
Subject:
Request for Additional information Regarding NRC Generic Letter 97-01, i
" Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head l
Penetrations" l
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References. (1) Letter from J.B. Hickman (U.S. NRC) to O.D. Kingsley (Comed),
" Generic Letter (GL) 97-01, " Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations," Responses for Braidwood l
Units 1 and 2, and Byron Units 1 and 2," dated September 2,1998.
(2) Letter from R.M. Krich (Comed) to U.S. NRC," Notification of Revised j
Date for Submittal of Additional Information Regarding NRC Generic '
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L Letter (GL) 97-01, " Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations"," dated. December 1,1998.
(3) Letter from D.J. Modeen (Nuclear Energy Institutrs) to G.C. Lainas (U.S. NRC), " Responses to NRC Requests for Additional Information on Generic Letter 97-01," dated December 11,1998.
l l-l (4) Letter from J.B, Hosmer (Comed) to U S. NR O, " Commonwealth Edison Company (Comed) Response to NR O Generic Letter 97-01, gf l
. Degradation of Control Rof Drive Mechanium Nozzle and Other
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Vessel Closure Head Penstrations"," dated April 1,1997.
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On September 2,1998, the Nuclear Regulatory Commission (NRC) i.ssued a request for additional information (Reference 1) tc Commonwealth Edison (Comed) Company to
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I support the review of Comed responses to Generic Letter (GL) 97-01, " Degradation of gb i
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CRDM/CEDM Nozzle and Other Vessel Closure Head Penetatiores." The request o
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9901200340 990115 7
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January 15,1999
. U.S. Nuclear Regulatory Commission i
Page 2 for additional information (RAI) letter asked that Comed provide additional information for each of our pressurized water reactor (PWR) designed plants within 90 days of the date of the RAI letter, i.e., by December 1,1998. As transmitted by Reference 2, we l
extended our plant response submittal date to January 15,1999, tc ailow additional time to incorporate the industry generic response into cur plant-spec % response for the Braidwood and Byron Stations.
Attached is our response to the RAl. Comed endorses the industry response to the RAl as submitted by the Nuclear Energy Institute (NEI) and transmitted by Reference 3, specifically the following Enclosures to the NEl submittal as they apply to the Braidwood I
and Byron Stations i, " Histogram of Reactor Pressure Vessel (RPV) Head Nozzle Assessments and Plant inspection Plans";
. Enclosure 2, " Responses to Generic NRC Requests for Additional Information";
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', " Westinghouse Owners Group (WOG) Responses to NRC Requests for Additionalinformation"; and i, " Description of the Electric Power Research Institute (EPRI) RPV Head Nozzle Primary Water Stress Corrosion Cracking (PWRSCC) Predictive Model."
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'If you have any questions conceming this correspondence, please contact Mr. J.A.
j.
Bauer at (630) 663-7287.
l Respectfully, l
l hor-R.M. Krich I
Vice President - Regulatory Services Attachment cc:
Regional Administrator-NRC Regionlli NRC Senior Resident inspector - Braidwnd Station NRC Senior Resident inspector - Byron Station j
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Attachment i
8 Request for Additional information Regarding NRC Generic Letter 97-01,
" Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations" l
BACKGROUND Commonwealth Edison (Comed) Company is a member of the V estinghouse Owners Group
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1 (WOG), and endorsed WCAP-14902, Revision 0, " Background laaterial for Response to NRC j
i Generic Letter 97-01: Reactor Vessel Closure Head Penetration Integrity for the Westinghouse j
Owners Group," transmitted as Attachment H by Reference 4, att applicable with respect to the
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i assessment of vessel head penetration (VHP) nozzles at Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2. Our personnel participated ir the development of the industry response, transmitted by Reference 3, to a request for additional information (RAl) made by the Nuclear Regulatory Coramission (NRC), transtr:cted by Reference 1. Comed contracted Dominion Engineering, Incorporated (DEI), through the Electric Power Research Institute (EPRI), to perform a probabilistic assessment of the potential for primary water stress corrosion cracking (PWSCC) in our VHP nozzles as a function of plant life. DEI developed and used the Control Rod Drive Mechanism (CRDM) Nozzle PWSCC Inspection and Repair Strategic Evaluation (CIRSE) Program for Comed plant specific probshilistic axessments.
REQUEST (1) Westinghouse Electric Corporation (WEC) and the WOG did not provide ;a description of the crack initiation and growth susceptibility model used '/or the assessment of WEC ver.sel head penetration (VHP) nozzles in plants endorsing WCAP-14902, Revisio, D. Provide a description of the crack initiation and growth susceptibility model L.r.,ed for the assessment of the VHP nozzles at your plants.
RESPONSE
A description of the crack initiation and grow'n susceptibility model used for the assessment of the VHP nozzles at Braidwood Station and Byron Station was submitted as,
transmitted by Reference 3.
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REQUEST (2) in WCAP-14902, Revision 0, WEC did not provide any conclusions as to what the probabilistic failure model would lead the WOG to conclude with respect to the assessment of PWSCC in WEC-designed vessel head penetration (VHP) noules.
With respect to the probabilistic susceptibility model (e.g., probabilistic failure model) provided in WCAP-14902, Revision 0:
(a) Provide the susceptibility ranking of your plant (s) as compiled from the crack initiation and growth analysis of the VHP noules for your plant (s) to that compiled for the other WOG member plants for which WCAP-14902, Revision 0, is applicable. Include the basis for establishing the ranking of your plant (s) relative to the others.
RESPCUSE The industry histogram and basis for ranking each plant has been provided to the NRC as Enclosure 1," Histogram of RPV Head Nozzle Assessments and Plant inspection Plans," transmitted by Reference 3. As can be seen from the industry histogram, c3raidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, are in the third assessment group. This group has over 15 effective full power years from January 1,1997, until the probability of having a crack at the allowable depth matches the DC Cook 2 probability of one 75% through-wall crack.
REQUEST (b) Describe how the probabilistic failure (crack initiation and growth) model used for the assessment of the VHP nozzles at your plant (s) was bench-marked, and provide a list and discussion of the standards the model was bench-marked against.
RESPONSE
Benchmarking for crack initiation is performed using a reference nozzle concept. After each plant inspection is completed, the vessel head and nozzles are analyzed using the EPRI model to determine the time to 10% probability of cracking for a reference nozzle with a surface hoop stress level of 60 ksi and an operating temperature of 600*
F which results in a 50% cumulative probability of the observed inspection results when corrections for differences in stress and temperature between the reference nozzle and the nozzles in the inspected plants are included. This information is then evaluated relative to the results of inspections for other plants to establish a time to 10% probability of crack initiation for each different group of nozzle materials.
t Crack growth is benchmarked using reported crack growth rates obtained from controlled laboratory tests and field inspections corrected for differences in temperature and crack tip stress intensity. Please refer to the EPRI methodology description in Enclosure 6 transmitted by Reference 3 for additional information on how the EPRI modelis benchmarked.
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REQUEST l
(c) Provide additional information regarding how the probabilistic failure (crack Initiation and growth) models for the assessment of the VHP nozzles at your plant (s) will be refined to allow the input of plant-specific inspection data into the model's analysis methodology.
RESPONSE
Plant-specific inspection data are factored into the EPRI model predictions le two ways.
- 1) As each plant inspection is completed, the vessel hea:i and nozzles are analyzed using the EPRI model to determine the time to 10% probability of cracking for a reference nozzle with a surface hoop stress level of 60 ksi and an operating temperature of 600* F which results in a 50% cumulative probability of the observed inspection results. These data are updated periodically and provided to users of the EPRI model software. If an inspection indicates a significant change in reference nozzle conditions, users are notified.
- 2) Once a plant has performed an inspection, the results of the plant-specific inspection, along with the results for other plants in the same nozzle material group, are used to establish a plant-specific reference for the future.
REQUEST (d) Describe how the variability in product forms, material specifications, and heat treatments used to fabricate each CRDM penetration nozzle at the WOG member utilities are addressed in the probabilistic crack initiation and growth models described or referenced in Topical Report No. WCAP-14902, Revision 0.
RESPONSE
The EPRI model time-to-crack-initiation predictions for a subject plant are based on the results of inspections at plants which most closely resemble the subject plant in terms of material product form, material specification, material supplier, material heat treatment, and vessel head fabricator. This approach avoids the need for maj'or corrections to reflect differences in material PWSCC susceptibility. Minor variations from nozzle-to-nozzle are accounted for statistically through the Weibull slope parameter, and by applying a triangular distribution to the reference time to 10%
probability of cracking. At the present time, EPRI considers that sufficient laboratory or field inspection data are not available to more precisely define the effect of product i
form, material specification, and heat treatment on the crack initiation rates. If proven correlations become available in the future, they will be included in the EPRI model.
All EPRI model crack growth predictions are based on application of a log-triangular distribution to the available laboratory and field data corrected for temperature and stress intensity.
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REQUEST (3) Table 1-2 in WCAP-14902, Revision 0, provides a summary of the key tasks in WEC's vessel head penetration nozzle assessment program. The tables indicate that the tasks for (1) Evaluation of PWSCC '4!tigation Methods, (2) Crack Growth Date '2nd Testing, and (3) Crack initiation characterization Studies hav6 not been mipleted and are stillin progress. In light of the fact that the probabilisth susceptibility models appear to be dependent in part on PWSCC crack initiation and growth estim etes, provide your best estimate when these tasks will be completed by WEC, and ctscribe how these activities relate to and will be used to update the probabilistic susceptibility assessment of VHP nozzles at your plant (s).
RESPONSE
The programs on crack growth testing and crack initiation have been Ossentially completed, and the program on mitigation is now underway and targeted for completion in mid-2000. These programs have thus far served to confirm the assumptions used in the original safety evaluations and models. As additionalinformation becomes available from the referenced testing, the models will be reviewed and updated as necessary, No major changes are anticipated.
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REQUEST i
(4) in the NEl letters of January 29,1998 and April 1,1998, NEl indicated that inspection plans have been developed for the VHP nonles at the Farley Unit 2 plant in the year 2002, and the Diablo Canyon Unit 2 plant in the year 2001, respectively. The staff has noted that although you have decided to apply an Wernate probabilistic model l
to the assessment of the VHP nonles at your plant (s), other WOG member l
licensees, including the Southern Nuclear Operating Company and the Pacific Gas and Electric Company, the respective licensees for the Farley units and the Diablo Canyon units, have selected to apply the susceptibility model described in WCAP-14901, Revision 0, to the assessment of VHP noules at their plants. The WOG's proposal to inspect the CRDM penetration no@as at Farley Unit 2 and Diablo Canyon Unit 2 sppears to be based on an composite assessment of the VHP nonles j
at all WOG member plants. Verify that such a composite ranking assessment has been applied to the evaluation of VHP nonles at your plant (s). If composite rankings j
of the VHP nonles at WOG member plants have been obtained from the composite l
results of the two models, justify why application of the alternate probabilistic susceptibility model being for the assessment of the VHP nonles at your plant (s) would yield the same comparable relative rankings as would application of the probabilistic susceptibility model used by the WOG member plants subscribing to the contents of WCAP-14901, Revision 0. Comment on the susceptibility rankings of j
the VHP nonles at your plant (s) relative to the susceptibility rankings of the VHP nonles at the Farley Unit 2 and Diablo Canyon Unit 2 plants.
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RESPONSE
i' The individual plant results are compared in the histogram shown in Enclosure 1 of the i
industry response transmitted by Reference 3. Enclosure 1 of the industry response also l
includes a summary table containing the three susceptibility categories shown in the histogram, which provide a comparison of the susceptibility for head penetration cracking i
of our Braidwood and Byron Stations relative to the Farley, Unit 2, and Diablo Canyon, j
Unit 2 plants. The Farley, Unit 2 plant is included in the highest calculated susceptibility i
(i.e., less than five effective full power years (EFPY)), and Diablo Canyon, Unit 2, is included in the group with an intermediate calculated susceptibility (i.e., five to fifteen EFPY). Braidwood and Byron Stations fall into the category with the lowest calculated i
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susceptibility (i.e., greater than 15 EFPY). Enclosure 1 of the industry response provides
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a description of the process used to compare individual plant analysis results to the D.C.
Cook, Unit 2 reference probability, i
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