ML042380592
| ML042380592 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/01/2004 |
| From: | Raghavan L NRC/NRR/DLPM/LPD3 |
| To: | Koehl D Nuclear Management Co |
| Chernoff, H, NRR/DLPM, 301-415-4018 | |
| References | |
| TAC MB9932, TAC MB9933 | |
| Download: ML042380592 (15) | |
Text
September 1, 2004 Dennis L. Koehl Site Vice President Point Beach Nuclear Plant Nuclear Management Company, LLC 6590 Nuclear Road Two Rivers, WI 54241
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2, RE: REQUEST FOR RELIEF FROM AMERICAN SOCIETY OF MECHANICAL ENGINEERS, BOILER AND PRESSURE VESSEL CODE, SECTION XI, REQUIRED EXAMINATION COVERAGE (TAC NOS. MB9932 AND MB9933)
Dear Mr. Koehl:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated July 3, 2003, the Nuclear Management Company, LLC (licensee), submitted a request for relief from certain requirements of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code),Section XI, 1986 Edition with no addenda. The licensee sought relief from the Code because the required 100 percent examination coverage could not be achieved for certain welds at the Point Beach Nuclear Plant (PBNP) Units 1 and 2.
The NRC staff has reviewed this request for relief. The NRC staffs safety evaluation is enclosed. Pursuant to Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.55a(g)(6)(i), relief is authorized on the basis that it was impractical for the licensee to meet the Code requirement of 100 percent examination coverage. Furthermore, reasonable assurance of structural integrity has been provided by the examinations that were performed by the licensee. Therefore, granting this relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
This relief is authorized for the third 10-year inservice inspection interval at PBNP, which ended June 30, 2003.
Sincerely,
/RA/
L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301
Enclosure:
Safety Evaluation cc w/encl: See next page
- Provided SE Input by memo NRR-028 OFFICE PDIII-1/PM PDIII-1/LA EMCB/SC*
OGC PDIII-1/SC NAME MCall HChernoff THarris TChan RHoefling LRaghavan DATE 08/19/04 08/19/04 5/04/04 &
08/25/04 08/31/04 09/01/04
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2, RE: REQUEST FOR RELIEF FROM AMERICAN SOCIETY OF MECHANICAL ENGINEERS, BOILER AND PRESSURE VESSEL CODE, SECTION XI, REQUIRED EXAMINATION COVERAGE (TAC NOS. MB9932 AND MB9933)
Dated:
DISTRIBUTION PUBLIC PDIII-1 Reading RidsNrrDlpmLpdiii-1 (LRaghavan)
RidsNrrPMHChernoff RidsNrrLATHarris RidsAcrsAcnwMailCenter RidsRgn3MailCenter (PLouden)
RidsOgcRp GGeorgiev GHammer DWeaver TChan GHill(4)
Point Beach Nuclear Plant, Units 1 and 2 cc:
Jonathan Rogoff, Esquire Vice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Mr. F. D. Kuester President & Chief Executive Officer WE Generation 231 West Michigan Street Milwaukee, WI 53201 Regulatory Affairs Manager Point Beach Nuclear Plant Nuclear Management Company, LLC 6610 Nuclear Road Two Rivers, WI 54241 Mr. Ken Duveneck Town Chairman Town of Two Creeks 13017 State Highway 42 Mishicot, WI 54228 Chairman Public Service Commission of Wisconsin P.O. Box 7854 Madison, WI 53707-7854 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Resident Inspectors Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road Two Rivers, WI 54241 Mr. Jeffery Kitsembel Electric Division Public Service Commission of Wisconsin P.O. Box 7854 Madison, WI 53707-7854 Nuclear Asset Manager Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, WI 53201 John Paul Cowan Executive Vice President & Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Douglas E. Cooper Senior Vice President - Group Operations Palisades Nuclear Plant Nuclear Management Company, LLC 27780 Blue Star Memorial Highway Covert, MI 49043 Site Director of Operations Nuclear Management Company, LLC 6610 Nuclear Road Two Rivers, WI 54241 January 2003
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF FROM AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE SECTION XI EXAMINATION COVERAGE REQUIREMENTS AT POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301
1.0 INTRODUCTION
Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g), specifies that inservice inspection (ISI) of nuclear power plant components shall be performed in accordance with the requirements of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code),Section XI, except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). The regulation at 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the U.S. Nuclear Regulatory Commission (NRC), if (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. Section 50.55a(g)(5)(iii) of 10 CFR states that if the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in Section 50.4, information to support the determinations.
By letter dated July 3, 2003, Nuclear Management Company, LLC (the licensee) submitted a request for relief pursuant to 10 CFR 50.55a(g)(5)(iii). The licensee sought relief from the requirements of ASME Code,Section XI (1986 Edition, no Addenda) because the required 100 percent examination coverage could not be achieved for certain welds at the Point Beach Nuclear Plant (PBNP), Units 1 and 2. The proposed relief was sought for the third 10-year ISI interval, which ended June 30, 2003. While the licensees submission of the stated letter was in conformance with the requirement set forth in Section 50.55a(g)(5)(iii), disposition of this type of relief request is done pursuant to Section 50.55a(g)(6)(i).
The licensee stated that PBNP used the 1974 Edition of Section XI for selection of Class 1 Category B-J piping welds. The 1974 Edition of Section XI requires that during each ISI interval, a different 25 percent of the population of ASME Code Class 1 piping welds be examined. Because PBNP was in its third 10-year interval, only 50 percent of the total Class 1 welds were available for selection and examination. For welds in all the other examination categories, PBNP used the Section XI required examination selection criteria. During previous intervals, most of the areas selected for examination were in low radiation areas and were relatively easy to access. As a result, for the third interval, a larger percentage of areas available for examination were in higher radiation areas or required additional support in order to gain access.
2.0 REGULATORY EVALUATION
The regulations at 10 CFR 50.55a require that ISI of ASME Code Class 1, 2, and 3 systems and their supports be performed in accordance with the applicable edition of the ASME Code and applicable addenda, except where alternatives have been authorized or relief has been requested by the licensee and granted by the Commission pursuant to paragraphs (a)(3)(i),
(a)(3)(ii), or (g)(6)(i) of 10 CFR 50.55a. In proposing alternatives or requesting relief, the licensee must demonstrate that: (1) the proposed alternatives provide an acceptable level of quality and safety; (2) compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for its facility. Section 50.55a allows the Commission to authorize alternatives and to grant relief from ASME Code requirements upon making the necessary findings.
The licensees proposed relief from certain requirements of Section XI of ASME Code-1986 with no Addenda, would apply only for the third 10-year ISI interval, which ended June 30, 2003.
The licensees evaluation and the NRC staffs findings with respect to granting or denying the ISI relief request are discussed below.
3.0 TECHNICAL EVALUATION
3.1 Licensees Evaluation 3.1.1 Component Identification Code Class: 1, 2 Examination Categories: See attached tables Item Numbers: See attached tables
==
Description:==
Limited Section XI Code examinations during Code required examinations 3.1.2 Code Requirements Third ISI Interval 1986 Edition of Section XI, no Addenda ASME Code Case N-460, Alternative Examination Coverage for Class 1 and 2 Welds Volumetric and surface examinations of welds and base material will be examined in accordance with the applicable Examination Category and Item Number.
3.1.3 Basis for Relief During the scheduled examination, examiners reported numerous instances of interference.
PBNP personnel evaluated each instance, and, when possible, alternative examination areas were selected to avoid having to examine restricted areas. This reduced the population of welds where a limited examination would be encountered. On welds where alternatives were not available, additional techniques were performed to increase coverage where possible.
These techniques included using steeper angle beam ultrasonic techniques, approaching the examination area from a different direction, or employing radiography. Additionally, when they allowed for additional coverage, minor grinding, removal of insulation (including asbestos insulation), and movement of supports or other components were done. In this manner, the examination area coverage was increased and the number of welds with limited examinations reduced to the extent practicable.
In order to gain additional access to the areas where limited examinations were encountered, major modification of components would be required. However, major modification of components is not a feasible approach nor is it required to obtain additional coverage. These modifications would include removal of structural supports not designed for removal and installation of temporary support structures to compensate for loss in load bearing capacity.
Such modifications would be costly and result in additional dose with marginal improvement in quality or safety.
PBNP performed system leakage tests in accordance with the pressure test requirements of Section XI, Examination Categories B-P and C-H. These pressure tests covered every component within the Code boundaries established by PBNP. There was no through wall leakage noted for any component during the third ISI interval. Where leakage at mechanical connections was noted, the connections were corrected in accordance with maintenance procedures.
3.1.4 Proposed Alternative PBNP proposed to use the examination volume or surface coverage obtained on the listed components during the third interval examinations in lieu of the Code required volumes and surfaces. The licensee stated that the coverage obtained meets the intent of ASME Code Section XI and provides an acceptable level of quality and safety.
3.1.5 Conclusion The examinations performed during the third ISI interval were performed to the extent possible.
Additional coverage was impractical, as modification of systems, structures, and components would have resulted in significant radiation exposure with a minimal increase in the level of quality and safety.
3.1.6 Period for Which Relief Is Requested Relief was requested for the third ISI interval at PBNP, which ended June 30, 2003.
3.2 Staff Evaluation By letter dated July 3, 2003, the licensee submitted a request for relief pursuant to 10 CFR 50.55a(g)(5)(iii) [Section 50.55a(g)(5)(iii)]. The licensee sought relief from the requirements of ASME Code,Section XI (1986 Edition, no Addenda) because the required 100 percent examination coverage could not be achieved for certain welds at the PBNP, Units 1 and 2. The proposed relief was sought for the third 10-year ISI interval, which ended June 30, 2003.
The licensee stated that PBNP used the 1974 Edition of Section XI for selection of Class 1, Category B-J piping weld, which specifies that for each ISI interval, a different 25 percent of the population of Class 1 piping welds shall be examined. Because PBNP was in its third 10-year interval, only 50 percent of the total Class 1 welds were available for selection and examination.
For welds in all the other examination categories, PBNP used the Section XI required examination selection criteria. During previous intervals, most of the areas selected for examination were in low radiation areas and were relatively easy to access. For the third interval, a larger percentage of areas available for examination were in higher radiation areas or required additional support in order to gain access.
The licensee provided detailed descriptions of the examined welds for which the Code coverage requirement could not be met. The licensee also provided descriptions of the coverage that was achieved for each of these welds. The attachment to this safety evaluation lists these welds; there were 22 welds in Unit 1 and 21 welds in Unit 2. The licensee examined the required areas to the extent practical using ultrasonic testing (UT) techniques with personnel qualified in accordance with Appendix VIll of Section XI and as implemented by the Performance Demonstration Initiative (PDI). Where possible and allowed by the applicable PDI Generic Procedure, additional angles were used to increase coverage. In many cases, no combination of ultrasonic angle beam examinations would cover the entire examination area.
Thus, radiography was performed on many welds to increase or obtain complete coverage. In addition, surface examination was performed to increase coverage; the maximum available surface area was examined. Where possible, additional insulation was removed, including asbestos insulation, and supports or other components were moved to provide additional coverage. In each case, the maximum feasible coverage was obtained.
The licensee also provided illustrations of the component configurations and UT techniques used in examining the welds located in areas with limited coverage. The licensee was able to obtain a substantial amount of the required coverage using radiography and/or surface examination techniques. Therefore, the licensee has performed the Code required weld examinations to the extent practical and in all cases has achieved greater than 31 percent weld coverage. The highest achieved coverage using a combination of examination methods was 100 percent.
The NRC staff evaluated the licensees submittal and found that performance of 100 percent weld examination coverage was impractical in this case and that the achieved coverage should have detected any existing patterns of degradation. This level of coverage, in conjunction with the fact that no indications of service flaws were identified as a result of the weld examinations, provides reasonable assurance of continued structural integrity for the examined welds.
4.0 CONCLUSION
The NRC staff concludes that it was impractical for the welds identified in the attachment to be examined to the extent required by the Code at PBNP. It is further concluded that reasonable assurance of structural integrity has been provided by the examinations that were performed by the licensee. Thus, granting relief pursuant to 10 CFR 50.55a(g)(6)(i) is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(i) for the third 10-year ISI interval at PBNP.
Attachments: Listing of Welds with Limited Examinations Principal Contributor: G. Georgiev Date:
ATTACHMENT ATTACHMENT Listing of Welds with Limited Examinations for the Third ISI Interval Point Beach Nuclear Plant Unit 1 Component Identification Cat./
Item no.
System or Component Limitation Description RPV-HFlange B-A B1.40 Reactor Pressure Vessel Exam is limited by welded attachments of head lift rig and flange configuration.
PZR-SprayNoz-IRS B-D B3.120 Pressurizer Inside Radius Section Exam is limited due to the nozzle configuration and permanent insulation support rings.
RHE-N1-IRS B-D B3.150 Regenerative Heat Exchanger Exam is limited due to the nozzle configuration.
RHE-N4-IRS B-D B3.150 Regenerative Heat Exchanger Exam is limited due to the nozzle configuration.
RC-34-MRCL-AI-03 B-F B5.130 Reactor Coolant Elbow is static cast stainless steel, has high attenuation, and has a limiting ~1/2 step.
Nozzle blend radius limits proper transducer contact.
RC-36-MRCL-AII-01 B-F B5.130 Reactor Coolant Elbow is static cast stainless steel, has high attenuation, and has a limiting ~1/2 step.
Nozzle blend radius limits proper transducer contact.
RC-34-MRCL-BI-03 B-F B5.130 Reactor Coolant Elbow is static cast stainless steel, has high attenuation, and has a limiting ~1/2 step.
Nozzle blend radius limits proper transducer contact.
Point Beach Nuclear Plant Unit 1 Component Identification Cat./
Item no.
System or Component Limitation Description RC-36-MRCL-BII-01 B-F B5.130 Reactor Coolant Elbow is static cast stainless steel, has high attenuation, and has a limiting ~1/2 step.
Nozzle blend radius limits proper transducer contact.
RPV Closure Nut B-G-1 B6.10 Reactor Pressure Vessel Middle 1/3 of the threaded area inside the nut is inaccessible.
RC-34-MRCL-BI-02 B-J B9.11 Reactor Coolant Elbow is static cast stainless steel and has high attenuation. Elbow has a limiting ~1/2 step.
Weld has a limiting ~1/2 step.
RC-36-MRCL-BII-02 B-J B9.11 Reactor Coolant Elbow is static cast stainless steel and has high attenuation. Elbow has a limiting ~1/2 step.
RC-36-MRCL-BII-06 B-J B9.11 Reactor Coolant Elbow is static cast stainless steel, has high attenuation, and has a limiting ~1/2 step. Pump configuration does not allow any examination.
RC-08-DR-1001-01 B-J B9.11 Reactor Coolant Exam limited due to the branch connection configuration.
SIS-10-SI-1003-19 B-J B9.11 Safety Injection Valve is cast stainless material and has high attenuation. Exam limited due to valve configuration.
RC-10-SI-1003-20 B-J B9.11 Reactor Coolant Valve is cast stainless material and has high attenuation. Exam limited due to valve configuration.
RC-10-SI-1003-21 B-J B9.11 Reactor Coolant Exam limited due to the branch connection configuration.
Point Beach Nuclear Plant Unit 1 Component Identification Cat./
Item no.
System or Component Limitation Description AC-10-SI-1001-19 B-J B9.11 Auxiliary Cooling (SI)
Valve is cast stainless material and has high attenuation. Exam limited due to valve configuration.
RC-10-SI-1001-21 B-J B9.11 Reactor Coolant Exam limited by three welded attachments.
AC-06-SI-1001-21 B-J B9.11 Auxiliary Cooling (SI)
Valve is cast stainless material and has high attenuation. Single side exam due to valve configuration.
AC-08-RHR-1006-02 C-F-1 C5.11B (Note 1)
Auxiliary Cooling (RHR)
Valve is cast stainless material and has high attenuation. Single side exam due to valve configuration.
SIS-04-SI-1001-11 C-F-1 C5.11B (Note 1)
Safety Injection Single side exam from elbow only due to flange configuration. Elbow inner radius prevents complete transducer contact.
1SI-850A-Welds C-G C6.20 Safety Injection Welds partially obstructed due to permanent restraints.
NOTE 1 - In accordance with Wisconsin Electric letter VPNPD-91-360 from J.J. Zach to NRC, dated October 16, 1991, Wisconsin Electric agreed to extend the criteria for selection of Class 2 piping welds to a wall thickness greater than 0.312 inches. The item number C5.11B was created to differentiate these welds from those required under the regular C-F-1 examination areas.
Point Beach Nuclear Plant Unit 2 Component Identification Cat./
Item no.
System or Component Limitation Description RPV-HFlange B-A B1.40 Reactor Pressure Vessel Exam limited by welded attachments of head lift rig and flange configuration.
PZR-SprayNoz-IRS B-D B3.120 Pressurizer Exam limited due to the nozzle configuration and permanent insulation support rings.
RHE-N1-IRS B-D B3.150 Regenerative Heat Exchanger Exam limited due to the nozzle configuration.
RHE-N4-IRS B-D B3.150 Regenerative Heat Exchanger Exam limited due to the nozzle configuration.
PZR-SurgeNoz-SE B-F B5.40 Pressurizer Single side exam due to the nozzle configuration.
RC-34-MRCL-AI-05 B-F B5.70 Reactor Coolant Exam limited by safe-end and nozzle geometries.
RC-34-MRCL-BI-05 B-F B5.70 Reactor Coolant Exam limited by safe-end and nozzle geometries.
RPV Closure Nut B-G-1 B6.10 Reactor Pressure Vessel Middle 1/3 of the threaded area inside the nut is inaccessible.
RC-34-MRCL-AI-04R1 B-J B9.11 Reactor Coolant Elbow is cast material and has high attenuation.
Both elbow and safe-end have a limiting ~1/2 step.
RC-36-MRCL-AII-01R1 B-J B9.11 Reactor Coolant Elbow is cast material and has high attenuation.
Both elbow and safe-end have a limiting ~1/2 step.
RC-36-MRCL-AII-03 B-J B9.11 Reactor Coolant Elbow is cast material, has high attenuation, and has a limiting ~1/8 step.
RC-34-MRCL-BI-04R1 B-J B9.11 Reactor Coolant Elbow is cast material and has high attenuation.
Both nozzle and safe-end have a limiting ~1/2 step.
Point Beach Nuclear Plant Unit 2 Component Identification Cat./
Item no.
System or Component Limitation Description RC-36-MRCL-BII-01R1 B-J B9.11 Reactor Coolant Elbow is cast material and has high attenuation.
Both elbow and safe-end have a limiting ~1/2 step.
RC-36-MRCL-BII-04 B-J B9.11 Reactor Coolant Elbow is cast material, has high attenuation, and has a limiting ~1/2 step.
Weld configuration prevents proper transducer contact.
RC-10-AC-2001-01 B-J B9.11 Reactor Coolant Exam limited due to the branch connection configuration.
RC-10-AC-2001-11 B-J B9.11 Reactor Coolant Valve is cast stainless material and has high attenuation. Exam limited due to valve configuration and welded attachment.
RC-06-SI-2002-26 B-J B9.11 Reactor Coolant Exam limited due to the welded attachment.
Elbow inner radius prevents complete transducer contact.
AC-10-RHR-2004-09 C-F-1 C5.11B (Note 1)
Auxiliary Cooling (RHR)
Exam limited due to the tee inside saddle configuration preventing complete transducer contact.
AC-08-RHR-2006-01 C-F-1 C5.11B (Note 1)
Auxiliary Cooling (RHR)
Exam limited due to the tee inside saddle configuration preventing complete transducer contact.
AC-08-RHR-2002-04 C-F-1 C5.11B (Note 1)
Auxiliary Cooling (RHR)
Exam limited due to the tee inside saddle configuration preventing complete transducer contact.
Point Beach Nuclear Plant Unit 2 Component Identification Cat./
Item no.
System or Component Limitation Description SIS-06-SI-2008-27 C-F-1 C5.11 Safety Injection Valve is cast stainless material and has high attenuation. Exam limited due to pipe restraint and valve configuration.
Note 1 - In accordance with Wisconsin Electric letter VPNPD-91-360 from J.J. Zach to NRC, dated October 16, 1991, Wisconsin Electric agreed to extend the criteria for selection of Class 2 piping welds to a wall thickness greater than 0.312 inches. The item number C5.11B was created to differentiate these welds from those required under the regular C-F-1 examination areas.