ML052550287
| ML052550287 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry, PROJ0704  |
| Issue date: | 09/06/2005 |
| From: | Matthew Mitchell NRC/NRR/DE/EMCB |
| To: | Eaton B Entergy Operations |
| Khanna M, DE/EMCB, 415-2150 | |
| References | |
| Download: ML052550287 (8) | |
Text
September 6, 2005 Bill Eaton, BWRVIP Chairman Entergy Operations, Inc.
Echelon One 1340 Echelon Parkway Jackson, MS 39213-8202
SUBJECT:
REVISED REQUEST FOR ADDITIONAL INFORMATION - BWRVIP-108:
BWR VESSEL AND INTERNALS PROJECT TECHNICAL BASIS FOR THE REDUCTION OF INSPECTION REQUIREMENTS FOR THE BOILING WATER REACTOR NOZZLE-TO-VESSEL SHELL WELDS AND NOZZLE BLEND RADII
Dear Mr. Eaton:
By letter dated November 25, 2002, you submitted for NRC staff review, Electric Power Research Institute (EPRI) proprietary report, BWRVIP-108, BWR Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii. By letter dated May 17, 2005, the staff issued a request for additional information (RAI) regarding its review of the BWRVIP-108 report. The staff has since determined that a revision to RAI-2 is needed. Therefore, the staff is issuing a revised RAI. However, please note that the only change that was made to the May 17, 2005, RAI is with respect to RAI-2. No changes were made to RAI-1 and RAI-3.
The BWRVIP-108 report presents the technical basis to reduce inspection requirements to boiling water reactor (BWR) pressure vessel (PV) nozzle-to-shell welds and nozzle blend radii.
Currently, BWR PV nozzle-to-shell welds and nozzle blend radii are inspected per American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI requirements, which require 100% inspection each 10-year interval. The purpose of the BWRVIP-108 report is to provide the technical basis for reducing the inspection requirements for the nozzle-to-shell welds and nozzle blend radii to 25% of the nozzles every 10 years. The 25% coverage refers to 25% of the nozzles for each nozzle type, e.g., 1 of 4 main steam nozzles would be inspected.
The staff has determined that additional information is needed to complete the review.
The staffs revised RAI regarding the BWRVIP-108 report is attached. It should be noted that these questions were raised after the staff had reviewed the Tennessee Valley Authority (TVA) and BWRVIPs responses to the staffs RAI, dated November 15, 2004, regarding TVAs relief request dated July 25, 2003, on the reduction of inspection frequency for the Browns Ferry, Units 2 and 3 nozzle-to-vessel shell welds and nozzle blend radii.
B. Eaton In order to complete the staffs review of the BWRVIP-108 report in an efficient and effective manner, your complete response to the attached RAI is required no later than six months from the date of this letter. If you cannot provide a complete response within six months, please contact Meena Khanna at (301) 415-2150 to discuss the withdrawal of the BWRVIP-108 report and its future resubmittal when you are prepared to respond to the RAI. In addition, if you have any other questions regarding the attached RAI, please contact Ms. Khanna.
Sincerely,
/(RA by M. Mitchell)/
Matthew A. Mitchell, Chief Vessels & Internals Integrity and Welding Section Materials and Chemical Engineering Branch Division of Engineering Office of Nuclear Reactor Regulation Project No. 704
Enclosure:
As stated cc: BWRVIP Service List
ML052550287 OFFICE EMCB:DE E
EMCB:DE E
EMCB:DE E
EMCB:DE E
EMCB:DE E
NAME MKhanna SSheng DNaujock TChan MMitchell DATE 08/04/2005 08/07/2005 08/05/2005 08/26/2005 09/06/2005
cc:
Tom Mulford, EPRI BWRVIP Integration Manager Raj Pathania, EPRI BWRVIP Mitigation Manager Ken Wolfe, EPRI BWRVIP Repair Manager Larry Steinert, EPRI BWRVIP Electric Power Research Institute P.O. Box 10412 3412 Hillview Ave.
Palo Alto, CA 94303 Robert Carter, EPRI BWRVIP Assessment Manager EPRI NDE Center P.O. Box 217097 1300 W. T. Harris Blvd.
Charlotte, NC 28221 George Inch, Technical Chairman BWRVIP Assessment Committee Constellation Nuclear Nine Mile Point Nuclear Station (M/S ESB-1) 348 Lake Road Lycoming, NY 13093 Denver Atwood, Chairman BWRVIP Repair Focus Group Southern Nuclear Operating Co.
Post Office Box 1295 40 Inverness Center Parkway (M/S B031)
Birmingham, AL 35242-4809 Jeff Goldstein, Technical Chairman BWRVIP Mitigation Committee Entergy Nuclear NE 440 Hamilton Ave. (M/S K-WPO-11c)
White Plains, NY 10601 Charles J. Wirtz, Chairman BWRVIP Inspection Focus Group FirstEnergy Corp.
Perry Nuclear Power Plant (M/S A250) 10 Center Road Perry, OH 44081 Amir Shahkarami BWRVIP Executive Oversight Committee Member Exelon Corporation 4300 Winfield Road Warrenville, IL 60555-4012 Robin Dyle, Technical Chairman BWRVIP Integration Committee Southern Nuclear Operating Co.
42 Inverness Center Parkway (M/S B234)
Birmingham, AL 35242-4809 Al Wrape, Executive Chairman BWRVIP Assessment Committee PPL Susquehanna, LLC 2 N. 9th St.
Allentown, PA 18101-1139 H. Lewis Sumner, Executive Chairman BWRVIP Mitigation Committee Vice President, Hatch Project Southern Nuclear Operating Co.
M/S BIN B051, P.O. BOX 1295 40 Inverness Center Parkway Birmingham, AL 35242-4809 Rick Libra BWRVIP Executive Oversight Committee Member DTE Energy 6400 N Dixie Hwy (M/S 280 OBA)
Newport, MI 48166-9726
ATTACHMENT REVISED REQUEST FOR ADDITIONAL INFORMATION REGARDING BWRVIP-108: BWR VESSEL AND INTERNALS PROJECT TECHNICAL BASIS FOR THE REDUCTION OF INSPECTION REQUIREMENTS FOR THE BOILING WATER REACTOR NOZZLE-TO-VESSEL SHELL WELDS AND NOZZLE BLEND RADII RAI-1. BWRVIP Response to NRC RAI 2-9b: In your probabilistic fracture mechanics (PFM) analysis, which supports the relief requests for Browns Ferry, Units 2 and 3 to use the reduced inspection frequency for the boiling water reactor (BWR) nozzle-to-vessel shell welds and nozzle blend radii, as recommended in the BWRVIP-108 report, the mean stress corrosion cracking (SCC) initiation time is assumed to be the same as that indicated in the BWRVIP-05 report, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations, which uses an initiation time of five times that of the curve that is applied to non-susceptible materials. The staffs independent analysis, as documented in the final Safety Evaluation (SE) on the BWRVIP-05 report, dated July 28, 1998, did not consider the reactor pressure vessel (RPV) cladding SCC initiation time (i.e., zero SCC initiation time). Therefore, the staff did not take a position on that issue in 1998. Since an initiation time of five times that of the curve, which is being applied to non-susceptible materials, is judgmental, the staff requests that the BWRVIP assess the impact of using the mean curve for cast austenitic stainless steel in BWR plants, (i.e., one times that of the BWRVIP-05 curve which is being applied to non-susceptible materials) on your PFM results. The staff requests that the BWRVIP continue the simulation until several failures have been observed. The staff would like to evaluate the probability of failure, based on simulated vessel failures, so that the PFM results could be adjusted, should it become necessary.
RAI-2. BWRVIP Response to NRC RAI 2-9e: The response states that the flaw size distribution is based on the pressure vessel research users facility (PVRUF) data, and the number of flaws in the weld is based the Marshall distribution. The flaws in the PVRUF data originate from steelmaking, hot rolling/forging RPV plates, welding the plates together, and heat treating the assembly. The PVRUF data does not include flaws that originate as a result of inservice/operating conditions.
The inservice inspections of the RPV nozzle-to-shell welds and RPV nozzle inner radii provide assurance that no degradation mechanism are diminishing the integrity of the RPV.
RPV NOZZLE INNER RADIUS EXAMINATIONS Early in 1980, operating conditions generated thermal fatigue cracks in BWR feedwater nozzle inner radii. The ultrasonic testing (UT) employed by the industry for the inner radii, at that time, was ineffective in detecting the thermal fatigue cracks.
The effectiveness of UT examinations of the nozzle inner radii steadily improved.
On June 5, 1998, the staff issued a safety evaluation on the Boiling Water Reactor Owners Group report GE-NE-523-A71-0594, Alternate BWR Feedwater Nozzle Inspection Requirements, which included demonstrating the detectability of a 1/4-inch notch and based the examination frequency on using techniques qualified to the objectives of Appendix VIII of Section XI of the American Society of Mechanical Engineers (ASME) Code. In December 1995, the ASME approved Code Case N-552,
2 Qualification for Nozzle Inside Radius Section from the Outside Surface, which has performance-based acceptance criteria. ASME Code Case N-552 was endorsed by the NRC in the Federal Register (64FR183) on September 22, 1999. The reliability and effectiveness of UT examinations of the RPV nozzle inner radii is assured by the performance-based qualification process of Supplement 5 to Appendix VIII of Section XI of the ASME Code or the alternative performance-based qualification process of Code Case N-552. Examinations of the RPV nozzle inner radius to the requirements of Appendix VIII, Supplement 5 became mandatory on November 22, 2002.
The staff is requesting a listing of the separate systems and nominal pipe sizes (the listing is the same as that required by Code Case N-702, Alternative Requirements for BWR Nozzle Inner Radius and Nozzle-to-Shell Welds) that apply to the RPV nozzle inner radii examination. For each item on the list, provide the total number of RPV nozzle inner radii (for all 35 BWR units). Also, for each item on the list, provide the total number of UT examinations performed using procedures that demonstrate the reliability of these examinations with respect to detection of cracks of various dimensions and locations, and which were performed by personnel that demonstrated their skills in reliably identifying these cracks. The establishment of UT reliability can be accomplished with statistically analyzed data from blind performance demonstrations, such as, Supplement 5 to Appendix VIII, spirit of Supplement 5 to Appendix VIII, and Code Case N-552.
RPV-TO-NOZZLE WELD EXAMINATION In the late 1970s, the Program for the Inspection of Steel Components (PISC) was organized to study the uncertainties in reliability and effectiveness of UT examinations of thick steel components, such as RPVs. In February 1983, the NRC issued Regulatory Guide 1.150, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice examinations, which adapted selected recommendations from the PISC program. Since 1995, personnel evaluated the applicability of Appendix VIII, Supplements 4 and 6 of ASME Code Section XI. Some licensees submitted relief requests to apply a combination of Supplements 4 and 6 as an alternative examination to the ASME Code-required (Section V) RPV-to-nozzle examinations.
By November 22, 2000, when NRC rulemaking made Supplements 4 and 6 mandatory, a sufficient number of personnel qualified to Supplements 4 and 6 were available to handle the industrys inservice inspection needs. The major differences between combined Supplement 4 and 6 RPV examinations and RPV-to-nozzle weld examinations are the limitations caused by the proximity of nozzle configurations to the RPV-to-nozzle welds. The ASME Code developed Appendix VIII, Supplement 7 to qualify personnel and procedures for RPV-to-nozzle weld examinations. On November 22, 2002, NRC rulemaking made compliance with Appendix VIII, Supplement 7 examinations mandatory.
The staff is requesting a listing of the separate systems and nominal pipe sizes (the listing is the same as that required by Code Case N-702) that apply to the RPV-to-nozzle weld examinations. For each item on the list, provide the total number of RPV-to-nozzle welds (for all 35 BWR units). Also, for each item on the list, provide the total number of UT examinations performed using procedures that demonstrated reliability of these examinations with respect to detection of cracks of various
3 dimensions and locations, and which were performed by personnel that demonstrated their skills in reliably identifying these cracks. The establishment of UT reliability can be accomplished with statistically analyzed data from blind performance demonstrations, such as, Appendix VIII, Supplement 7 qualified personnel and procedures and staff-approved (via the relief request process) Supplements 4 and 6 qualified personnel and procedures.
Since many of these examinations achieved less than essentially 100% coverage, identify on the list the number of RPV-to-nozzle weld examinations with less than essentially 100% coverage in ranges, i.e., 0 to 25%, 25% to 50%, 50% to 75%, 75% to less than 90%, and 90% and greater.
RAI-3. BWRVIP Response to NRC RAI 2-10: In the RAI, the staff describes a PFM analysis using the PFM parameters that were employed in the staffs independent PFM evaluations related to the BWRVIP-05 review (Interim SE dated August 14, 1997; SE dated July 28, 1998) and the Palisades pressurized thermal shock (PTS) review (SE dated April 12, 1995) as the worst case study. The staffs wording of worst case did not appropriately characterize the nature of the study that assesses the sensitivity of key PFM parameters. The staff requests that the BWRVIP assess the sensitivity of the following PFM parameters:
C Parameter (b), the number of flaws per nozzle for nozzle blend radius, which is based on the SE for the Palisades PTS review, where a flaw density for plates was assumed to be 1/10 of that of the weld; C
Parameters (c) and (d), the standard deviation for fracture toughness and upper-shelf fracture toughness of 15% of the mean values, which are those used by the staff in both the BWRVIP-05 review and the Palisades PTS review; C
Parameter (e), the SCC growth rate, approximately 10 times your rate, which is based on the staffs BWRVIP-05 review; C
Parameter (f), threshold stress intensity factor of 10 ksi/in, which is based on current industry effort in characterizing another type of SCC.