ML021970012
| ML021970012 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/11/2002 |
| From: | Richard Laufer NRC/NRR/DLPM/LPD1 |
| To: | Skolds J Exelon Generation Co, Exelon Nuclear |
| Tam P, NRR/DLPM, 415-1451 | |
| References | |
| TAC MB2940 | |
| Download: ML021970012 (8) | |
Text
July 11, 2002 Mr. John L. Skolds, President and Chief Nuclear Officer Exelon Nuclear Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
OYSTER CREEK NUCLEAR GENERATING STATION - RELIEF REQUEST R17, REVISION 1, EXAMINATION COVERAGE FOR REACTOR PRESSURE VESSEL AXIAL WELDS (TAC NO. MB2940)
Dear Mr. Skolds:
By letter dated September 14, 2001, as supplemented June 3, 2002, AmerGen Energy Company, LLC (AmerGen), submitted Relief Request R17, Revision 1. The submittals requested relief for Oyster Creek Nuclear Generating Station (OCNGS) from certain American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, requirements, regarding volumetric examination coverage requirements for reactor pressure vessel (RPV) welds.
The 1986 Edition of the ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item B1.10, requires examination of all welds in the first inspection interval and one beltline region weld in successive inspection intervals. However, Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g)(6)(ii)(A)(2) concurrently requires all licensees to augment their RPV examinations by implementing once, as part of the inservice inspection interval in effect on September 8, 1992, the examination requirements for RPV shell welds specified in item B1.10 of Examination Category B-A, Pressure Retaining Welds in Reactor Vessel (i.e., inspection of essentially 100% of the volume of the vessel axial welds). AmerGen requested Nuclear Regulatory Commission (NRC) authorization of its proposed alternative to the requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2).
The NRC staff reviewed the referenced submittals and has set forth details of its findings in the enclosed safety evaluation. Based on its review, the NRC staff concludes that the approximately 60% composite coverage proposed by AmerGen provides an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes the proposed alternative for the third 10-year inservice inspection interval at OCNGS.
J. L. Skolds If you have any questions regarding this relief, please call Mr. Peter Tam, NRC Project Manager, at (301) 415-1451.
Sincerely,
/RA by SRichards for RLaufer/
Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-219
Enclosure:
Safety Evaluation cc w/encl: See next page
J. L. Skolds If you have any questions regarding this relief, please call Mr. Peter Tam, NRC Project Manager, at (301) 415-1451.
Sincerely,
/RA by SRichards for RLaufer/
Richard J. Laufer, Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-219
Enclosure:
Safety Evaluation cc w/encl: See next page DISTRIBUTION:
PUBLIC OGC SRichards PD1-1 R/F GHill (2) RLaufer TSteingass PTam SLittle ACRS JRogge, RI TChan BPlatchek, RI TBergman, RI, EDO Accession Number: ML021970012 *SE provided by memo dated 6/18/02 OFFICE PD1-1/PM PD1-1/LA EMCB/SC OGC PDI-1/SC NAME PTam SLittle TChan* RHoefling SRichards for RLaufer**
DATE 7/2/02 7/2/02 6/18/02 7/10/02 7/11/02
- S. Richards signed and concurred for R. Laufer.
OFFICIAL RECORD COPY
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION INSERVICE INSPECTION PROGRAM, RELIEF REQUEST R17, REVISION 1 OYSTER CREEK NUCLEAR GENERATING STATION AMERGEN ENERGY COMPANY, LLC DOCKET NO. 50-219
1.0 INTRODUCTION
The Inservice Inspection (ISI) of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Class 1, Class 2, and Class 3, components is to be performed in accordance with Section XI of the ASME Code and applicable edition and addenda as required by Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). In addition, 10 CFR 50.55a(a)(3) states in part that alternatives to the requirements of paragraph (g) may be used, when authorized by the Nuclear Regulatory Commission (NRC), if the applicant demonstrates that: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Pursuant to 10 CFR 50.55a(g)(4), ASME Code, Class 1, 2, and 3, components (including supports) will meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The ISI code of record for Oyster Creek Nuclear Generating Stations (OCNGS) third 10-year ISI interval is the 1986 Edition of the ASME Code.
By letter dated September 13, 2000, the NRC staff approved relief request R17 for the OCNGS reactor pressure vessel (RPV) Code, Category B-A, volumetric examination coverage limitations. The limitation percentages were based on an internal vessel accessibility study of the RPV prior to performing the subject examinations. By letter dated September 14, 2001, as supplemented June 3, 2002, AmerGen Energy Company, LLC (the licensee), submitted Relief Request R17, Revision 1, for the same OCNGS volumetric examination coverage requirements for the same RPV welds, Code Category B-A, based on actual percentages calculated after completion of the volumetric examinations. These actual percentages are substantially less than the percentages projected from the accessibility study.
Enclosure
2.0 NRC STAFF EVALUATION 2.1 Code Requirements for which Relief is Requested ASME Code Section XI, Table IWB-2500-1, Category B-A, Item B1.12 requires examination of all welds in the first inspection interval and one beltline region weld in the successive inspection intervals. However, 10 CFR 50.55a(g)(6)(ii)(A)(2) states that all licensees shall augment their reactor vessel examinations by implementing the examination requirements for RPV welds in Item B1.10 of Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, in Table IWB-2500-1 of subsection IWB of the 1989 Edition of ASME Code,Section XI, Division 1, subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(A)(2). For the purposes of this augmented examination, essentially 100%, as used in Table IWB-2500-1, means more than 90% of the examination volume for each weld.
2.2 Licensees Basis for Relief The licensee determined that conformance with the volumetric coverage requirements is impractical. The OCNGS reactor, a BWR-2, was designed and built before ASME Code,Section XI was developed and access for inspections became a design requirement. The licensee indicated that there is little external access to the outside diameter of the RPV axial shell welds due to inadequate clearance between the bioshield wall and vessel insulation. This access limitation forced the licensee to perform an internal surface directed ultrasonic inspection of the axial shell welds using the General Electric (GE) GERIS-2000 inspection system.
The welds where limited examination volume was achieved and the reason for the limitation is listed in Table 1 below:
Table 1 Weld Examination Limitations Code Volume Examined Identification NR02 2-563A* None 100.0%
NR02 2-563B* Steam Dryer Support Lug 92.8%
NR02 2-563C Main Steam Nozzle Plug 80.4%
NR02 2-563D Feedwater Sparger, Manipulator Scan Limits 62.3%
NR02 2-563E Feedwater Sparger, Manipulator Scan Limits 59.7%
NR02 2-563F Feedwater Sparger, Manipulator Scan Limits 59.7%
NR02 2-564A Feedwater Sparger, Manipulator Scan Limits 0.0 NR02 2-564B* None 100.0%
NR02 2-564C Shroud Repair Tie Rod 39.7%
NR02 2-564D Recirculation Outlet Nozzle 56.1%
NR02 2-564E Shroud Repair Tie Rod 74.7%
NR02 2-564F Feedwater Sparger Brackets Interference with Scanner 0.0
- Meets Code requirements for examination volume The licensee stated that it had submitted Relief Request R17 based on the estimated coverage that would be obtained. This estimated coverage was based on GEs assessment of weld locations and clearances based on drawings. The actual coverages obtained after completion of the examinations are listed in Table 2 below:
Table 2 Weld Weld Weld Estimated Actual Identification Configuration Length Coverage Coverage NR02 2-563A* Longitudinal, Upper Shell @ 15o AZ 132.6" 100.0% 100.0%
NR02 2-563B* Longitudinal, Upper Shell @ 135o AZ 132.6" 99.2% 92.8%
NR02 2-563C Longitudinal, Upper Shell @ 255o AZ 132.6" 99.4% 80.4%
NR02 2-563D Longitudinal, Int. Upper Shell @ 330o AZ 132.6" 65.3% 62.3%
o NR02 2-563E Longitudinal, Int. Upper Shell @ 90 AZ 132.6" 65.3% 59.7%
NR02 2-563F Longitudinal, Int. Upper Shell @ 210o AZ 132.6" 62.6% 59.7%
NR02 2-564A Longitudinal, Lower Int. Shell @ 219o AZ 133.6" 93.0% 0.0 NR02 2-564B* Longitudinal, Lower Int. Shell @ 339o AZ 133.6" 93.0% 100.0%
o NR02 2-564C Longitudinal, Lower Int. Shell @ 99 AZ 133.6" 94.1% 39.7%
NR02 2-564D Longitudinal, Lower Shell @ 258o AZ 83.4" 55.1% 56.1%
NR02 2-564E Longitudinal, Lower Shell @ 18o AZ 131.6" 76.0% 74.7%
NR02 2-564F Longitudinal, Lower Shell @ 138o AZ 131.6" 76.0% 0.0
- Meets Code requirements for examination volume Welds 2-563D, 2-563E, 2-563F, and 2-564E, coverages were slightly less than the estimated coverages by 3 to 6%. Weld 2-563C was estimated to achieve the ASME Code coverage but was limited to 80.4% actual due to the main steam line plug which was unanticipated during the accessibility study. Weld 2-564A was not located at the azimuth appearing on the drawing, but was located behind a tie rod, which is six degrees counter clockwise from the position shown on the drawing. The same tie rod offset configuration existed for weld 2-564C which reduced examination coverage to 39.7%. Weld 2-564F is located directly below a feedwater nozzle where the nozzle-to-sparger configuration makes it impossible to position the transducer array flush with the weld.
The licensee stated that the NRC staff acceptance of the original relief request was based upon an estimated 82% total coverage with 100% coverage of the axial welds in the beltline region.
The actual was 60% total coverage with 57% coverage of the axial welds in the beltline region.
Since the actual coverage was significantly less than estimated, a failure probability was
performed using the VIPER code developed by Structural Integrity Associates, Inc. (SIA). The evaluation was based on the methodology presented in Electric Power Research Institute (EPRI) Report, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations, BWRVIP-05. Based on the evaluation, the inspections performed were effective in inspecting regions that account for essentially all of the vessel fracture failure risk. The examination coverage resulted in a small failure probability of 2.5x10-12/reactor-year risk which the licensee argued is an acceptable level of quality and safety.
2.3 Evaluation The 1986 Edition of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item B1.10, requires examination of all welds in the first inspection interval and one beltline region weld in successive inspection intervals. However, 10 CFR 50.55a(g)(6)(ii)(A)(2) concurrently requires all licensees to augment their RPV examinations by implementing once, as part of the ISI interval in effect on September 8, 1992, the examination requirements for RPV shell welds specified in Item B1.10 of Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, in Table IWB-2500-1 of subsection IWB of the 1989 Edition of Section XI of the ASME Code, subject to the conditions specified in 10 CFR 50.55a(g)(6)(ii)(A)(3) and (4). The licensee is requesting NRC authorization of its alternative to the requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2), which requires an augmented examination of essentially 100% of the volume of the vessel axial welds. Three welds in the lower intermediate shell course were approximately six degrees counter clockwise from the location specified on the drawings. Two welds could not be examined because of the shroud repair tie rod and feedwater sparger bracket interferences.
The licensee stated that it had performed a failure probability of the welds in the beltline region using the VIPER code developed by SIA. The evaluation by SIA was performed to assess the effect on the probability of fracture due to the 57% composite coverage performed on the vessel axial welds. The evaluation was based on the methodology presented in EPRI Report, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations, BWRVIP-05. The licensee stated the inspections performed were effective in inspecting regions that account for essentially all of the vessel fracture failure risk based on a very small failure probability of 2.5x10-12/reactor-year as a result of the 57% composite beltline examination coverage.
The failure frequency value is determined by the product of the conditional failure probability and a low temperature overpressurization transient probability. The licensee stated in its letter dated June 3, 2002, that OCNGS specific weld chemistry values were used in the evaluation and these values are significantly lower than the bounding values used in the BWRVIP-05 calculations. Secondly, the failure probabilities in the BWRVIP-05 were determined using bounding values for weld chemistry and RPV geometry. The BWRVIP-05 analysis used an RPV thickness of 5.25 inches and a diameter of 225.2 inches. This compares with a thickness of over 7 inches and a diameter of 213 inches for the OCNGS vessel. The licensee stated that these differences result in a significant reduction in the hoop stress compared to the BWRVIP-05 evaluation, which significantly reduces the failure probability for OCNGS. Though the staff did not review the probability risk assessment in detail, the greater thickness and lower diameter of the OCNGS vessel provide a reasonable basis for the calculated low failure probability of 2.5x10-12/reactor-year. This value is sufficiently below the RPV failure frequency due to failure of the limiting axial welds in the boiling water reactor fleet at 5x10-6/reactor-year,
which is consistent with the guidelines of Regulatory Guide 1.154, and provides reasonable assurance of a low failure probability due to the limited examination coverage.
Finally, though the actual coverage of each weld was significantly lower than the projected coverage, the NRC staff believes that the actual coverage obtained would identify any pattern of degradation, particularly in the high fluence region of the vessel wall. The high fluence region is the area that experiences the highest neutron bombardment and would be the first region to display any embrittlement issues caused by irradiation-assisted cracking. The NRC staff concludes that the approximately 60% composite coverage provides an acceptable level of quality and safety.
3.0 CONCLUSION
Based on the discussion above, the NRC staff concludes that the alternative will provide an acceptable level of quality and safety, and is in compliance with 10 CFR 50.55a(a)(3)(i).
Therefore, pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5), the NRC staff authorizes the proposed alternative for the third 10-year inservice inspection interval for OCNGS, identified as Relief Request R17, Revision 1.
Principal Contributor: T. Steingass Date: July 11, 2002
Oyster Creek Nuclear Generating Station cc:
Chief Operating Officer Mayor of Lacey Township Exelon Generation Company, LLC 818 West Lacey Road 4300 Winfield Road Forked River, NJ 08731 Warrenville, IL 60555 Senior Resident Inspector Senior Vice President - Nuclear Services U.S. Nuclear Regulatory Commission Exelon Generation Company, LLC P.O. Box 445 4300 Winfield Road Forked River, NJ 08731 Warrenville, IL 60555 Director - Licensing Vice President - Mid-Atlantic Operations Exelon Generation Company, LLC Support Correspondence Control Desk Exelon Generation Company, LLC P.O. Box 160 200 Exelon Way, KSA 3-N Kennett Square, PA 19348 Kennett Square, PA 19348 Oyster Creek Generating Station Plant Senior Vice President - Manager Mid Atlantic Regional Operating Group AmerGen Energy Company, LLC Exelon Generation Company, LLC P.O. Box 388 200 Exelon Way, KSA 3-N Forked River, NJ 08731 Kennett Square, PA 19348 Regulatory Assurance Manager Kevin P. Gallen, Esquire Oyster Creek Nuclear Generating Station Morgan, Lewis, & Bockius LLP AmerGen Energy Company, LLC 1800 M Street, NW P.O. Box 388 Washington, DC 20036-5869 Forked River, NJ 08731 Kent Tosch, Chief Vice President, General Counsel and New Jersey Department of Secretary Environmental Protection Exelon Generation Company, LLC Bureau of Nuclear Engineering 300 Exelon Way CN 415 Kennett Square, PA 19348 Trenton, NJ 08625 J. Rogge, Region I Vice President - U.S. Nuclear Regulatory Commission Licensing and Regulatory Affairs 475 Allendale Road Exelon Generation Company, LLC King of Prussia, PA 19406-1415 4300 Winfield Road Warrenville, IL 60555 Manager Licensing - Oyster Creek and Three Mile Island Site Vice President Exelon Generation Company, LLC Oyster Creek Nuclear Generating Station Nuclear Group Headquarters AmerGen Energy Company, LLC Correspondence Control PO Box 388 P.O. Box 160 Forked River, NJ 08731 Kennett Square, PA 19348 H. J. Miller Correspondence Control Desk Regional Administrator, Region I Exelon Generation Company, LLC U.S. Nuclear Regulatory Commission 200 Exelon Way, KSA 1-N-1 475 Allendale Road Kennett Square, PA 19348 King of Prussia, PA 19406-1415