ML12178A192
| ML12178A192 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/25/2012 |
| From: | Thadani M C Plant Licensing Branch 1 |
| To: | Harding T Constellation Energy Nuclear Group |
| Thadani M | |
| References | |
| ISI-07, ISI-08, TAC ME8000, TAC ME8001 | |
| Download: ML12178A192 (3) | |
Text
From:Thadani, MohanTo:Harding Jr, ThomasCc:Alley, David; Sheng, SimonSubject:Gina: TAC NOS. ME8000 and ME8001-Relief Requests ISI-07 and ISI-08-Regarding BMIDate:Monday, June 25, 2012 10:56:00 AMAttachments:image001.pngTom:
The NRC staff is reviewing the requests ISI-07 and ISI-08 for relief pursuant to 10 CFR50.55a, regarding reactor pressure vessel closure head penetration nozzle examinationsfor R.E. Ginna, during its fifth Inservice Inspection interval. The NRC staff has identified aneed for following additional information required to facilitate the continuation of review ofthe relief requests.
Please provide your response to the following request promptly to facilitate the completionof the NRC staff's review in time to support your requested schedule for NRC response. REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF FROM REQUIREMENTS FOR EXAMINATION OF REACTOR VESSEL HEAD PENETRATION NOZZLES R. E. GINNA NUCLEAR POWER PLANT, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
RAI-1 The May 24, 2012, submittal did not provide information on axial stresses for the controlrod drive mechanism (CRDM) penetration nozzles on the basis that the hoop stresses arelimiting. As indicated in Pacific Northwest National Laboratory (PNNL) Report 17763,"Final Report - Inspection Limit Confirmation for Upper Head Penetration NozzleCracking," dated August of 2008, this basis is not always true. Case 12 of Figure 9, Case61 of Figure 11, and Cases 87, 88, 89 of Figure 13 of the PNNL report all indicated thatthe axial stresses are more limiting than the hoop stresses. When you study the abovequoted figures in the PNNL report, please note that, for axial stresses, the longer distancefrom the triple point to where the stress drops to 20 ski means that a longer portion of thepenetration nozzle is at stresses greater than 20 ski.
Please provide additional plant-specific information on axial stresses to substantiate yourclaim that the hoop stresses are limiting.
RAI-2 The licensee used a deterministic fracture mechanics analysis to justify the alternativeexamination coverage for penetration nozzle number 35. However, the proposed coverageof 0.2 inch is not sufficient to tolerate the uncertainties in the calculated residual stressesand the determination of the assumed crack length. The staff also has a concern with theproposed crack growth rate for Alloy 690 CRDM material (discussed in RAI-3). Hence, theNRC staff requests the licensee provide the basis for not performing a surface examinationof the lower portion of each penetration nozzle necessary to meet the inspectionrequirements of 10 CFR 50.55a(g)(6)(ii)(D) and consider changing the regulatory basis ofthe relief request from 10 CFR 50.55a(a)(3)(i) to10 CFR 50.55a(a)(3)(ii). Up to date, allsimilar relief requests were granted under 10 CFR 50.55a(a)(3)(ii), because the NRC stafffinds insufficient basis to grant relief under 10 CFR 50.55a(a)(3)(i), considering thatsurface examinations could be performed on each penetration nozzle to meet the currentinspection requirements. RAI-3 The licensee used a crack growth rate with an improvement factor of 100 over Alloy 600for Alloy 690 materials in its fracture mechanics analyses. The May 24, 2012, submittalstates that NUREG/CR-7103, Volume 2, "Pacific Northwest National LaboratoryInvestigation of Stress Corrosion Cracking [SCC] in Nickel-Base Alloys," indicated similarcrack growth results except for high levels of cold work. However, the NUREG states,"Representative [Alloy] 690 CRDM plant materials were studied to investigate the SCCcrack-growth response in the as-received TT [thermally treated] and several modifiedconditions." To ensure that the plant-specific Alloy 690 material for the CRDMs did notreceive significant cold work which is "representative" according to the NUREG, pleaseprovide your estimate of the cold work for the Ginna CRDM penetrations based on thefabrication and installation record of the Ginna CRDM penetrations to support that theGinna CRDM penetrations did not receive significant cold work.
RAI-4 Since the May 24, 2012, submittal did not specify the examination frequency, the staffassumed that the examination frequency of ASME Code Case N-729-1 will be followed.Please provide EFPYs operated so far for the Ginna CRDM penetration nozzles andspecify the time for the next examination based on (1) whether prior examinations haveidentified flaws and (2) whether the amount of cold work requested in RAI-3 allows you toconsider the plant-specific Alloy 690 material as PWSCC resistant. Please note that ifinsignificant cold work cannot be demonstrated for the Ginna CRDM penetration nozzles,the nozzles' classification of Item No. B4.40 of the American Society of MechanicalEngineers Boiler and Pressure Vessel Code Code Case N-729-1, "Alternative ExaminationRequirements for PWR [pressurized water reactor] Reactor Vessel Upper Heads WithNozzles Having Pressure-Retaining Partial-Penetration Welds," may not be justified.
If there are any questions, please contact me. Mohan C ThadaniSenior Project ManagerPlant Licensing Branch IVDivision of Operating Reactor Licensing Office of Nuclear Reactor Regulation(301) 415-1476 Mohan.Thadani@nrc.gov