ML20212H949
| ML20212H949 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 06/23/1999 |
| From: | Thadani M NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| TAC-MA4886, NUDOCS 9906280227 | |
| Download: ML20212H949 (4) | |
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UNITED STATES g
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20565-0001 49*****
,o June 23, 1999 MEMORANDUM TOf!DnekEFile FROM:
Mohan C. Thadani, Senior Project Manager, Section 2 Project Directorate i Division of Licensing Project Management Office of Nuclear Reactor Regulation
SUBJECT:
PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 - DRAFT 1
REQUEST FOR ADDITIONAL INFORMATION REGARDING REQUEST TO REDUCE RE-INSPECTION FREQUENCY OF CERTAIN CORE SHROUD WELDS (TAC NO. MA4886 )
a The PECO Energy Company's request to reduce re-inspection frequency of certain core i
shroud welds for Peach Bottom Atomic Power Station, Unit 3, was discussed by the Nuclear Regulatory Commission (NRC) staff with the licensee's staff during a conference call held on June 21,1999. As a follow-up to the conference call, the attached draft request for additional information (RAl) was transmitted by facsimile on June 23,1999, to Mr. John Hufnagel of the licensee's staff. Review of the RAI would allow the licensee to determine and agree upon a schedule to respond to the RAl. This memorandum and the attachment do not convey a formal
/I request for information or represent an NRC staff position.
f Docket Nos. 50-277 and 50-278
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Attachment:
_ Request for Additional f./
Information Uti
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P Juns 23, 1999 MEMORANDUM TO: Docket File FROM:
Mohan C. Thadani, Senior Project Manager, Section 2 ORIG SIGNED BY:
Project Directorate i Division of Licensing Project Management Office of Nuclear Reactor Regulation
SUBJECT:
PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 - DRAFT REQUEST FOR ADDITIONAL INFORMATION REGARDING REQUEST TO REDUCE RE-INSPECTION FREQUENCY OF CERTAIN CORE SHROUD WELDS (TAC NO. MA4886 )
The PECO Energy Company's request to reduce re-inspection frequency of certain core shroud welds for Peach Bottom Atomic Power Station, Unit 3, was discussed by the Nuclear Regulatory Commission (NRC) staff with the licensee's staff during a conference call held on June 21,1999. As a follow-up to the conference call, the attached draft request for additional information (RAl) was transmitted by facsimile on June 23,1999, to Mr. John Hufnagel of the licensee's staff. Review of the RAI would allow the licensee to determine and agree upon a rehedule to respond to the RAl. This memorandum and the attachment do not convey a formal request for information or represent an NRC staff position.
Docket Nos. 50-277 and 50-278
Attachment:
Request for Additional Information DISTRIBUTION Docket File PUBLIC PDI-2 R/F E. Adensam J. Clifford M. Thadani DOCUMENT NAME:_
Dl-2\\PEACHBOTTOM\\RAIA4886.WPD b
OFFICE l
NAME MThadani/rsi f
DATE 6 /D99 OFFICIAL RECORD COPY
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REQUEST FOR ADDITIONAL INFORMATION
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~ REGARDING PECO ENERGY COMPANY'S REQUEST FOR REDUCING THE RE INSPECTION FREQUENCIES OF CERTAIN CORE SHROUD WELDS AT PEACH BOTTOM ATOMIC _ POWER STATION. UNIT 3 1.
You stated in your flaw evaluation that UT Inspection uncertainty factors were applied to all identified indications, as well as to uninspected regions. However, in note 2 to Tables 2,4,5, and 6, it is stated that for near side detection, the factor is 0.000" for 45 shear.
The staff notes that in your earlier flaw evaluation provided in your letter dated November 3,1995, an NDE uncertainty of 0.4 inches plus half a degree was applied to the end of each flaw. Please provide details regarding how and where the UT uncertainty factors including the zero uncertainty factor used in the flaw evaluation were determined.
Were mock-up assemblies used in determining the UT uncertainty for each weld? For welds H3 and H4, provide the following information regarding (a) the UT uncertainty factor applied to each flaw and the uninspected areas and (b) the inspection technique and the transducers used in measuring the flaw length.
~2.
In your flaw evaluation, the final length of each Indication was determined by considering the UT uncertainty, crack growth, and potential combination with other indications based on the proximity rule. To facilitate the staff's review, please provide detailed information pertaining to those intermediate steps taken in determining the final flaw sizes and distribution which were used in your linear elastic fracture mechanics (LEFM) flaw evaluation including the new initiation study for welds H3 and H4:
(i) original size of each detected flaw including the un-inspected areas (ii) adjusted length of each flaw after addition of UT uncertainty (iii) adjusted length of each flaw after addition of crack growth (iv) length of combined flaws based on p;oximity rule (v) the final distribution and sizes of flaws used for evaluation 3.
In Appendix A for limit load evaluation, the ANSC (licensee's software) was used for determining net section collapse of arbitrarily thinned cylinder. Please describe in detail the methodology used in ANSC program versus the methodology recommended in BWRVIP-07. Has this program and coding been validated and its results benchmarked against the calculations from the Nuclear Regulatory Commission (NRC) approved program for limit load evaluation or against hand calculations?
4.
In Appendix B for LEFM evaluation, the interaction effect from adjacent flaws was determined by using the methods presented by Rooke and Cartwright in " Compendium of Stress intensity Factor," The Hillington Press,1976. You indicated that this method is less conservative than the guidelines presented in BWRVIP-01,"BWR Core Shroud inspection and Flaw Evaluation Guideline, Revision 2," October,1996. However, you did not provide a comparison of the results calculated from the two methods. Please calculate the Attachment
,,1 2-interaction effect based on BWRVIP-01 guidelines and evaluate the resulting stress intensity factor (K) for welds H3 and H4. You also stated that the interaction effect was determined by considering a model of three unequal length collinear flaws, however, the diagram shown in Figure B-1 depicted a flaw with two adjacent flaws of equal length and distance. Please explain the discrepancy and discuss its impact to the evaluation of K.
5.
In Appendices B and D, the stress factor"A"is based on a method of averaging. The staff notes that when the crack pattern is not symmetric to X and Y axes of the core shroud, the bending axis of the core shroud will shift from the center of the core shroud to a new position. Consequently, the stresses due to the bending moment and pressure differential may be several times larger than those presented in Table 1. Please address the staff's concern by (1) examining the sources and natures for the bending moment and the pressure differential stresses, and (2) considering the effect of the lack of symmetry of the crack pattern to the stresses listed in Table 1.
6.
In Appendix D, the effect of welding residual stresses on K was considered for pariial through-wall flaw, however, the magnitude of the residual stress and its profile was not discussed in detail. The staff notes that weld H3 is a weld joint for the core shroud cylinder to the ring segment. Please discuss and provide supporting data regarding how the residual stresses in weld H3 were determined.
7.
In Appendix B for the LEFM evaluation of through-wall flaws, the effect of residual stresses on the stress intensity factor (K) was not discussed. Please give reasons why the residual stresses were not considered in the evaluation of K. What would be the value of K for welds H3 and H4 when the residual stresses resulting from welding, cold spring, weld repair and other sources are considared.
E, The acceptable methodologies for limit load and LEFM evaluations are provided in BWRVIP-07. Please provide a line-by-line comparison of the methodologies used in your evaluations to that provided in BWRVIP-07. When BWRVIP-07 guidelines are not followed, provide the reasons for the deviations and discuss the conservatism of the alternative method used in the evaluation.
9.
'You stated that at the end of Cycle 13, welds H3 and H4 will have fluence of approximately 6.0 x 10 ' n/cm and 1.7 x 10 ' n/cm', respectively. Please describe in detail how the 2
2 fluence was calculated and its range of uncertainties. Was the methodology used in the fluence calculation previously approved by the NRC? Have the analytical procedures or coding been validated and its results benchmarked against measured fluence?
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