NRC 2007-0020, 10 CFR 50.55a Requests, Relief Requests RR-18, RR-19 and RR-20 Associated with Examination of the Reactor Pressure Vessels Fourth Ten-Year Inservice Inspection Program Interval

10 CFR 50.55a Requests, Relief Requests RR-18, RR-19 and RR-20 Associated with Examination of the Reactor Pressure Vessels Fourth Ten-Year Inservice Inspection Program Interval

ML070990077
Person / Time
Site:	Point Beach
Issue date:	04/06/2007
From:	Koehl D Nuclear Management Co
To:	Document Control Desk, NRC/NRR/ADRO
References
NRC 2007-0020
Download: ML070990077 (12)

Text

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Committed to Nucleer Ex-Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC April 6, 2007 NRC 2007-0020 10 CFR 50.55a U.S. Nuclear Regulatory Commission AlTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 10 CFR 50.55a Requests Relief Requests RR-18, RR-19 and RR-20 Associated With Examination of the Reactor Pressure Vessels Fourth Ten-Year lnservice Inspection Proaram Interval This letter requests that the Nuclear Regulatory Commission (NRC) grant the Nuclear Management Company, LLC (NMC) relief from and authorize alternatives to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV Code),Section XI, 1998 Edition with 2000 Addenda for the following examinations of the Point Beach Nuclear Plant (PBNP) Units 1 and 2 reactor pressure vessels (RPVs):

RR-18, Table IWB-2500-1, Category B-K, ltem Number BlO.10, Welded Attachments for Vessels Piping, Pumps, and Valves, for examination of the welded attachments on the outside surface of the RPVs.

RR-19, Table IWF-2500-1, Category F-A, ltem Number F1.40, Supports, for the examination of certain supports on the RPV.

RR-20, Table IWB-2500-1, Category B-A, ltem Number 81-30, to allow use of ASME Section XI, Appendix Vlll for examination of the RPV Upper Vessel Shell-to-Flange Welds.

Relief is being requested in accordance with the requirements of 10 CFR 50.55a(g)(5)(iii). Compliance with the specified requirements is impractical without a compensating increase in the level of quality or safety for RR-18 and RR-19.

Performance of the specified examinations in RR-18 and RR-I9 would require modification to the reactor pressure vessel (RPV) biological shield walls and support system to gain access to all of the areas required to be examined. RR-20 is an alternative examination, using enhanced ultrasonic examination techniques which provide an acceptable level of quality and safety.

6610 Nuclear Road Two Rivers, Wisconsin 54241-9516 Telephone: 920.755.2321

Document Control Desk Page 2 NMC requests approval of these relief requests by March 15, 2008. The Unit 2 Refueling 29 (U2R29) outage is scheduled to commence in April 2008. RPV examinations will be conducted during this refueling outage. The requested duration of these relief requests is for the remainder of the Fourth Ten-Year Inservice Inspection Program, which is scheduled to end on June 30, 2012, for both PBNP units.

Summarv of Commitments This submittal contains no new commitments or revisions to existing commitments.

Dennis L. Koehl /

Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC Enclosures cc: Regional Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Mr. Mike Verhagan, Department of Commerce, State of Wisconsin

ENCLOSURE I RELIEF REQUEST RR-18 REQUEST FOR RELIEF FOR EXAMINATION OF WELDED ATTACHMENTS ON THE REACTOR PRESSURE VESSEL ASME Code Components Affected Point Beach Nuclear Plant (PBNP) Units 1 and 2 Welded attachments on the outside surface of the reactor pressure vessel (RPV) for both units.

Applicable Code Edition and Addenda

The inservice inspection (ISI) program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 1998 Edition with 2000 Addenda.

Applicable Code Requirements ASME Section XI, Examination Category B-K, Item B1O.10 requires a surface examination of essentially 100% of the length of a weld for all welded attachments which meet the following conditions:

a) The attachment is on the outside surface of a pressure retaining component; b) The attachment provides component support as defined in NF-1110; c) The attachment weld joins the attachment either directly to the surface of the component or to an integrally cast or forged attachment to the component, and d) The attachment weld is full penetration fillet, or partial penetration, continuous or intermittent.

"Essentially 1OOOh,"as clarified by ASME Code Case N-460, is any examination which achieves greater than 90% coverage of the examination volume or surface area, as applicable.

lmpracticabilitv of Compliance The RPV welded attachments are identified in the PBNP IS1 Plan as welds RPV-MK-16 and RPV-MK-17, which are located at vessel azimuths 88.5" and 268.5", respectively.

The configuration of the above identified welded attachments is similar to ASME Section XI, Figure IWB-2500-15. Access to the RPV inlet and outlet pipinglnozzles is provided by the removal of sand plugs (blocks) located directly above each nozzle-to-reactor coolant piping connection weld. However, due to the placement Page 1 of 4

of the welded attachments inside of permanent insulation and the biological shield walllreactor vessel support structure, the welds are inaccessible for examination by either the surface or the visual examination method. In addition, due to its placement within a small annulus region between the biological shield wall and the reactor pressure vessel, a majority of the vessel's support system is inaccessible for visual (VT-3) examination. The configuration of these welded attachments is shown in Figure 2.

During the First and Second Ten-Year Intervals, the welded attachments were examined utilizing ultrasonic (UT) techniques applied from the inside surface (ID) using automated examination equipment during the 10-year RPV inservice inspection (ISI).

These examinations revealed no recordable flaw-like indications for either unit. During the Third Ten-Year Interval, ASME Section XI, 1986 Edition, did not require the examination of these welds.

Burden Caused by Compliance In order to gain access to either the welded attachments or to a majority (>go%) of the area of the RPV support structure, significant modifications of the currently installed biological shield wall, vessel support system, and permanent insulation would be required. These modifications would entail significant personnel radiation exposure without a compensating increase in the level of quality and safety.

Proposed Alternative and Basis for Use NMC proposes that a UT examination of the welded attachments be performed utilizing ASME Section XI, Appendix VIII, "Performance Demonstrationfor Ultrasonic Examination Systems," Supplement 6, "Qualification Requirements for Reactor Vessel Welds Other Than CladIBase Metal Interface." The examinations will be performed with proceduresltechniquesconsistent with the requirements of Appendix VIII, Supplement 6 during the regularly scheduled Fourth Interval Ten-Year RPV examinations. The examination area would be limited to the surface shown on Figure IWB-2500-15 B-C, to a depth of one inch (1") into the material.

Duration of Proposed Alternative The duration of the proposed alternative is for the Fourth Ten-Year IS1 interval, which ends on June 30,2012.

Amendments 63 and 68 to Facility Operating Licenses Nos. DPR-24 and DPR-27, dated February 17, 1983, "Safety Evaluation on Request for Relief from Inservice Inspection Requirements, Wisconsin Electric power Company, Point Beach Nuclear Plant Units 1 & 2, Docket Nos. 50-266 & 50-301" Page 2 of 4

References ASME Section XI, 1998 Edition with Addenda through 2000 Performance Demonstration Initiative (PDI) Program Description, Revision 4 Figure I Proposed Ultrasonic Examination Area for Welded Attachments RPV-MK-16 and RPV-MK-17 (Based on Figure IWB-2005-15)

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Figure 2 PBNP Unit 1 RPV General Outline (Typical of Both Units)

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RELIEF REQUEST RR-I9 REQUEST FOR RELIEF FOR EXAMINATION OF CERTAIN SUPPORTS ON THE REACTOR PRESSURE VESSEL ASME Code Components Affected PBNP, Units Iand 2 Supports for the RPV for both units.

Applicable Code Edition and Addenda

ASME Section XI, 1998 Edition with 2000 Addenda Applicable Code Requirements Examination Category F-A, Item F1.40 requires a Visual, VT-3 for 100% of the supports other than piping supports (Class 1,2,3, and MC).

Impracticality of Compliance Access to the RPV inlet and outlet pipinglnozzles is provided by the removal of sand plugs (blocks) located directly above each nozzle-to-reactor coolant piping connection weld. However, due to its placement within a small annulus region between the biological shield wall and the RPV, a majority of the RPV support system is inaccessible for visual (VT-3) examination.

Burden Caused by Compliance In order to gain access to a majority of the area (>go%) of the RPV support structure, significant modifications of the currently installed biological shield wall, vessel support system, and permanent insulation would be required. These modifications would entail significant personnel radiation exposure without a compensating increase in the level of quality and safety.

Proposed Alternative and Basis for Use NMC proposes that a visual examination (VT-3) of the portions of the RPV support system accessible through the four reactor coolant nozzlelpiping access holes, as well as the portions available from below the RPV, be performed in accordance with the currently approved Inservice Inspection Program Fourth Interval Class 1, 2 and 3 Plan.

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Duration of Proposed Alternative The duration of the proposed alternative is for the Fourth Ten-Year IS1 Interval, which ends on June 30,2012.

Precedents None References ASME Section XI, 1998 Edition with Addenda through 2000 Page 2 of 2

ENCLOSURE 3 RELIEF REQUEST RR-20 REQUEST FOR RELIEF TO USE ASME SECTION XI, APPENDIX Vlll AND PERFORMANCE DEMONSTRATION INITIATIVE (PDI) FOR REACTOR VESSEL FLANGE ASME Code Components Affected Point Beach Nuclear Plant (PBNP) Units Iand 2 Reactor Pressure Vessel (RPV) Upper Vessel Shell-to-Flange Welds.

Applicable Code Edition and Addenda

The IS1 program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 1998 Edition with 2000 Addenda.

Applicable Code Requirements ASME Code Class I Reactor Pressure Vessel (RPV) Upper Vessel Shell-to-Flange Welds, Table IWB-2500-1 Category B-A, Item Number B1.30 requires an ultrasonic examination of the RPV shell-to-flange weld. In accordance with ASME Section XI, paragraph IWA-2232, "Ultrasonic examinations shall be conducted in accordance with Appendix I." Further, in accordance with Appendix I, Paragraph 1-2110(b), "Ultrasonic examination of reactor vessel-to-flange welds, closure head-to-flange welds, and integral attachment welds shall be conducted in accordance with Article 4 of Section V, except that altemative examination beam angles may be used."

Reason for Request

Performance of ultrasonic (UT) examinations which have been qualified through the ASME Section XI, Appendix VIII/Perforrnance Demonstration Initiative (PDI) process provides a superior examination compared to ASME Section V, Article 4 examinations. The proposed altemative represents the best techniques, procedures, and qualifications available to perform UT of RPV welds.

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Proposed Alternative and Basis for Use The listed weld is the only circumferential shell welds in the RPVs that are not examined in accordance with the requirements of ASME Section XI, Appendix VIII, as mandated in 10 CFR 50.55a with the issuance of the rule change contained in Federal Register 64 FR 51370, dated September 22,1999. This rule change mandated the use of ASME Section XI, Appendix VIII, Supplements 4 and 6 for the conduct of all other RPV weld examinations.

ASME Section V, Article 4, describes the required techniques to be used for the ultrasonic test (UT) of welds in ferritic pressure vessels with wall thicknesses greater than 2 inches. The techniques were first published in ASME Section V in the 1974 Edition, Summer 1975 Addenda. The calibration techniques, recording criteria and flaw sizing methods are based upon the use of a distance-amplitude-correction curve (DAC) derived from machined reflectors in a basic calibration block.

UT performed in accordance with Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations,"

Revision 1 and Section V, Article 4, used recording thresholds of 50 percent DAC for the outer 75 percent of the required examination volume and 20 percent DAC from the cladlbase metal interface to the inner 25 percent region of the examination volume. Indications detected in the designated exam volume portions, with amplitudes below these thresholds, were therefore not required to be recorded. Use of the Appendix Vlll (PDI) processes would enhance the quality of the examination results reported. The detection sensitivity is more conservative and the procedure requires the examiner to evaluate all indications determined to be flaws regardless of their associated amplitude. The recording thresholds in Section V, Article 4, the guidelines of Regulatory Guide 1.150, Revision 1 are generic and do not take into consideration such factors as flaw orientation, which can influence the amplitude of UT responses.

EPRl Report NP-6273, "Accuracy of Ultrasonic Flaw Sizing Techniques for Reactor Pressure Vessels," dated March 1989, established that UT flaw sizing techniques based on tip diffraction are the most accurate. The qualified prescriptive-based UT procedures of ASME Section V, Article 4 have been applied in a controlled process with mockups of RPVs which contained real flaws and the results statistimlly analyzed according to the screening criteria in Appendix Vlll of ASME Section XI. The results show that the procedures in Section V, Article 4, are less effective in detecting flaws than procedures qualified in accordance with Appendix Vlll as administered by the PDI processes.

Appendix VIIIIPDI qualification procedures use the tip diffraction techniques for flaw sizing. The proposed alternative Appendix VIIIIPDI UT methodology uses analysis tools based upon echo dynamic motion and tip diffraction criteria which have been validated. This methodology is considered more sensitive and accurate than the Section V, Article 4 processes.

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UT performed in accordance with the Section V, Article 4 processes requires the use of beam angles of 0°, 45", and 60" with recording criteria that precipitates equipment changes. Having to perform these process changes results in increased radiation exposure for examination personnel. Using these examination processes, personnel must examine the weld manually from the seal surface during reactor pressure vessel (RPV) head lift activities to achieve the maximum coverage of the weld(s). Compliance with the specific ASME Section XI, Appendix I requirements for the RPV circumferential shell-to-flange weld when the data is obtained using a less technically advanced process, results in an examination that does not provide a compensating increase in quality and safety for the higher personnel exposures incurred.

For future RPV shell-to-flange weld examinations using PDI-qualified techniques, PBNP does not anticipate any less coverage than the required minimum of 90 percent of coverage. However, if limitations are encountered during the conduct of the examinations, individual relief requests will be submitted, as needed.

Procedures, equipment, and personnel qualified via the Appendix VIII, Supplements 4 and 6 PDI programs have been shown to have a high probability of detection of flaws and are generally considered superior to the techniques employed earlier for RPV examinations. Accordingly, approval of this alternative evaluation process is requested pursuant to 10 CFR 50.55a(a)(3)(i).

Duration of Proposed Alternative The duration of the proposed alternative is for the Fourth Ten-Year IS1 interval, which ends on June 30,2012.

Precedents

1. Duke Energy Corporation submittal for Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Unit 2, and Oconee Nuclear Station, Unit 3, dated July 14, 2004, "Request for Relief for Use of an Alternate to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, for Reactor Vessel Examinations RR-04-GO-002" Page 3 of 4

2. NRC Safety Evaluation dated October 20, 2004, for Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Unit 2, and Oconee Nuclear Station, Unit 3, dated July 14, 2004, "Request for Relief for Use of an Alternate to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, for Reactor Vessel Examinations RR-04-GO-002 (TAC Nos. MC3804, MC3805, MC3807, and MC3810)" (ML420040261)

3. Tennessee Valley Authority Submittal dated February 23, 2005, for Browns Feny Units 1, 2 and 3; Sequoyah Nuclear Plant, Units 1 and 2; and Watts Bar Unit 1, "Relief Request to Use ASME Section XI, Appendix Vlll and Performance Demonstration lnitiative (PDI) for Reactor Vessel Flange Welds - PDI-4" (ML050590046)

4. NRC Safety Evaluation dated August 2, 2005, for Browns Feny Units 1,2 and 3; Sequoyah Nuclear Plant Units 1 and 2; and Watts Bar Nuclear Plant Unit 1, "Inservice Inspection Program Relief Request PDI-4 (TAC Nos. MC6232, MC6233, MC6234, MC6235, MC6236, and MC6237)"

(ML051730487)

ASME Section XI, 1998 Edition with Addenda through 2000 Performance Demonstration lnitiative (PDI) Program Description, Revision 4 Page 4 of 4