L-MT-07-071, CFR 50.55a Request No. 15: Relief from Impractical Examination Coverage Requirements Pursuant to 10 CFR 50.55a(g)(5)(iii) for the Fourth Ten-Year Inservice Inspection Interval

CFR 50.55a Request No. 15: Relief from Impractical Examination Coverage Requirements Pursuant to 10 CFR 50.55a(g)(5)(iii) for the Fourth Ten-Year Inservice Inspection Interval

ML072710119
Person / Time
Site:	Monticello
Issue date:	09/26/2007
From:	O'Connor T Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-07-071
Download: ML072710119 (18)

Text

Monticello Nuclear Generatinu Plant Operated by Nuclear Management Company, LLC Committed to Nudear E x ~ l e ~

September 26,2007 L-MT-07-071 10 CFR 50.55a(g)(5)(iii)

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket 50-263 License No. DPR-22 10 CFR 50.55a Request No. 15: Relief from Impractical Examination Coveraqe Requirements Pursuant to 10 CFR 50.55a(q)(5)(iii) for the Fourth Ten-Year lnservice lnspection Interval Pursuant to 10 CFR 50.55a(g)(5)(iii), the Nuclear Management Company, LLC (NMC) requests relief from certain examination coverage requirements imposed by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice lnspection of Nuclear Power Plant Components," for the Monticello Nuclear Generating Plant (MNGP). This 10 CFR 50.55a request is for weld examinations, performed during the 2007 refueling outage, where the required coverage of "essentially 100 percent" could not be obtained when examined to the extent practical. The basis for the 10 CFR 50.55a request is that compliance with the specified requirements is impractical due to plant design. The details of the 10 CFR 50.55a request are enclosed.

NMC is submitting this request for the Fourth Ten-Year lnservice lnspection Interval scheduled to end on May 31, 2012. If you have any questions or require additional information, please contact Lynne Gunderson at 715-377-3430.

This letter contains no new commitments and makes no revisions to existing commitments.

Site vice President, Monticello Nuclear Generating Plant Nuclear Management Company, LLC Enclosures (3) cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce 2807 West County Road 75 Monticello, Minnesota 55362-9637 Telephone: 763.295.5151 Fax: 763.295.1454

ENCLOSURE I 10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY

1. ASME Code Component(s) Affected Components affected are American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code (Code),Section XI, Class 1, Reactor Vessel Nozzle-to-Vessel welds specified below and in-detail in Table A:

Recirculation Inlet Nozzle N-2B Weld - N-2B NV Recirculation Inlet Nozzle N-2G Weld - N-2G NV Feedwater l nlet Nozzle N-4A Weld - N-4A NV Reactor Head Spare Nozzle N-6A Weld - N-6A NV Capped Control Rod Drive (CRD) Return Nozzle N-9 Weld - N-9 NV

2. Applicable ASME Section XI Code Edition and Addenda The applicable ASME Section XI Code for the Monticello Nuclear Generating Plant (MNGP), Fourth Ten-Year lnservice lnspection (ISI) Interval is the 1995 Edition with the 1996 Addenda.

3. Applicable Code Requirement

ASME Class 1 Nozzle-to-Vessel welds are subject to the examination requirements of Subsection IWB Table IWB-2500-1, as shown below, and 10 CFR 50.55a(b)(2)(xv)(G). The welds are required to be examined once within the Fourth Ten-Year Interval:

Code Class: 1

References:

IWB-2500, Table IWB-2500-1 Examination Category: B-D Item Number: B3.90

Description:

Nozzle-to-VesselWelds Component Numbers: See Section 1 and Table A System: Reactor Vessel Examination Method: Volumetric - Ultrasonic Testing (UT)

Examination Volume: Figure IWB-2500-7(b)

In August 2005, the Nuclear Regulatory Commission (NRC) issued Regulatory Guide (RG) 1.147, Revision 14, lnservice lnspection Code Case Acceptability, ASME Section XI, Division 1 (Reference 1). In RG 1.147, the NRC identifies the ASME Code Cases that they have determined to be acceptable alternatives to applicable parts of Section XI, and that these Code Cases may be used by licensees without requesting authorization from the NRC provided that they are used with any identified limitations or modifications. RG 1.147, Table 1 lists the Page 1 of 7

ENCLOSURE 1 10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50m55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY following two Code Cases as acceptable to the NRC for use by a licensee with no identified limitations or modifications: 1) Code Case N-460 (Reference 2),

and 2) Code Case N-613-1 (Reference 3).

Code Case N-460 states in part, "when the entire examination volume or area cannot be examined due to interference by another component or part geometry, a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10 percent."

NRC Information Notice (IN) 98-42 (Reference 4) termed a reduction in coverage of less than 10 percent to be "essentially 100 percent." IN 98-42 states in part, "The NRC has adopted and further refined the definition of 'essentially 100 percent' to mean 'greater than 90 percent'...has been applied to all examinations of welds or other areas required by ASME Section XI."

Code Case N-613-1 provides an alternative examination volume that includes the width of the weld plus one-half inch of adjacent base metal on each side of the widest part of the weld. In comparison, the examination volume required by the Figure IWB-2500-7(b) includes the width of the weld plus the adjacent base metal on each side of the widest part of the weld equal to one-half of the vessel shell wall thickness.

4. lmpracticalitv of Compliance Construction Permit CPPR-31 was obtained for the MNGP in 1967. The MNGP systems and components were designed and fabricated before the examination requirements of ASME Section XI were formalized and published. Therefore, MNGP was not specifically designed to meet the requirements of ASME Section XI and full compliance is not feasible or practical within the limits of the current plant design.

10 CFR 50.55a recognizes the limitations to inservice inspection of components in accordance with Section XI of the ASME Code that are imposed due to early plants' design and construction, as follows:

10 CFR 50.55a(g)(l): For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971, components (including supports) must meet the requirements of paragraphs (g)(4) and (5) of this section to the extent practical.

Page 2 of 7

ENCLOSURE I 10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50n55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY 10 CFR 50.55a(g)(4): Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and pre-service examination requirements, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code ... to the extent practical within the limitations of design, geometry and materials of construction of the components.

10 CFR 50.55a(g)(5)(iii): 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 § 50.4, information to support the determinations.

The inspection limitations on the subject components are due to inherent nozzle design geometric contours (see Table A).

A description of the examination methodology used to provide the maximum obtainable coverage is provided in Section 6 of this request. This methodology is based on ASME Section XI, Appendix Vlll qualification and was applied to the extent practical within the design constraints of the components. Enclosure 3 provides cross-sectional diagrams of the subject welds showing the geometric contour of the component design in relation to the welds and the coverage obtained within the examination volume requirements of Code Case N-613-1, Figure 2.

5. Burden Caused by Compliance Compliance with the examination coverage requirements of ASME Section XI would require modification, redesign, or replacement of components where geometry is inherent to the component design.

6. Proposed Alternative and Basis for Use Proposed Alternative In accordance with 10 CFR 50.55a(g)(5)(iii), relief is requested for the components listed in Table A on the basis that the required examination coverage of "essentially 100 percent" is impractical due to physical obstructions and the limitations imposed by design, geometry and materials of construction.

Page 3 of 7

ENCLOSURE I 10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50m55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY Nuclear Management Company (NMC) performed qualified examinations that achieved the maximum, practical amount of coverage obtainable within the limitations imposed by the design of the components. Additionally, as Class 1 examination Category B-P components, a VT-2 examination is performed on the subject components of the Reactor Coolant Pressure Boundary (RCPB) during system pressure tests each refueling outage. This was completed during the 2007 refueling outage and no evidence of leakage was identified for these components.

Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), NMC requests relief from the requirements of ASME Section XI Table IWB-2500-1, Category B-D, Item B3.90, and proposes to utilize these completed exams as acceptable alternatives that provide reasonable assurance of continued structural integrity.

Basis for Use The NMC Nondestructive Examination (NDE) procedures incorporate inspection techniques qualified under Appendix Vlll of the ASME Section XI Code by the Performance Demonstration Initiative (PDI) for examination of the subject nozzle-to-vessel welds, and allow the examination volume to meet the provisions of alternative requirements (i.e., Code Case N-613-1).

The examinations were performed using a manual contact method from the nozzle outside blend radius and vessel surfaces. Coverage was obtained by following the scan parameters designated within NMC NDE procedures and as defined by MNGP specific Electric Power Research Institute (EPRI) computer modeling reports (References 5 and 6) for each nozzle configuration and angle.

It should be noted that that the scans defined by the EPRI report are only applicable to the inner 15 percent of the weld volume when scanning in the parallel direction.

The refracted longitudinal wave mode of propagation was applied for all the radial scans of the exam volume, and to the outer 85 percent of the exam volume for parallel scans. The shear wave mode of propagation was applied for each of the transducer and wedge combinations required for the remaining inner 15 percent of the parallel scan exam volume.

The subject components received the required examination(s) to the extent

. practical within the limited access of the component design. One hundred (100) percent coverage was obtained for the inner 15 percent of the examination volume. The examination limitations for the subject components were encountered within the outer 85 percent of the examination volume. For the examinations conducted, satisfactory results were achieved, and no evidence of unacceptable flaws was detected with the inspection techniques.

Page 4 of 7

ENCLOSURE I 10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50a55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY Due to the design of these welds it was not feasible to effectively perform a volumetric examination of "essentially 100 percent" of the required volume. The nozzle-to-vessel welds are accessible from the vessel plate side of the weld and are examined to the extent practical, but there are no qualified examinations to obtain coverage of the excluded areas within the outer 85 percent of the examination volume due to the nozzle forging curvature.

Additional coverage for the limited areas was not achievable or practical, based on the latest qualified ultrasonic technology, nor by other considered examinations methods, such as radiography. NMC has concluded that if significant degradation existed in the subject welds, it would have been identified by the examinations performed.

Additionally, as Class 1 examination category B-P components, VT-2 examinations were performed on the subject components in association with the Reactor Coolant Pressure Boundary system pressure test performed during the 2007 refueling outage. No evidence of leakage was identified during this system test.

The materials for the subject components are A508 CI II nozzle forgings welded to A533 CI I vessel shell plate. A review of operating experience within the nuclear industry did not reveal any instances of cracking in this location and type of weldment.

The MNGP reactor vessel water chemistry is controlled in accordance with the 2004 revision to the BWR Water Chemistry Guidelines (Reference 7). Also a hydrogen water chemistry system is used to reduce the oxidizing environment in the reactor coolant. These additional measures provide added assurance against the initiation of cracking or corrosion from the inside surface of the reactor vessel. An inerted primary containment environment during operation provides assurance of corrosion protection on the outside surface of the reactor vessel.

The provisions described above as an alternative to the code requirement will continue to provide reasonable assurance of the structural integrity of the subject welds. The examinations were completed to the extent practical and evidenced no unacceptable flaws present. VT-2 examinations performed on the subject components during system pressure testing each refueling outage (in accordance with examination Category B-P) provide continued assurance that the structural integrity of the subject components is maintained. Additionally, the MNGP Water Chemistry Program and inerted primary containment environment provide added measures of protection for the component materials. Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), NMC requests relief from the ASME Section XI examination requirements for the subject nozzle-to-vessel welds.

Page 5 of 7

ENCLOSURE I 10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50m55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY

7. Duration of Proposed Alternative NMC requests the granting of this relief for the Fourth Ten-Year lnservice lnspection Interval of the lnservice lnspection Program for the MNGP that is scheduled to end on May 31,2012.

8. Precedents The NRC has granted relief for other nozzle-to-vessel shell welds at the MNGP, most recently for the current Fourth Ten-Year lnservice lnspection Interval (Reference 8). Also, the NRC has granted relief for the Quad Cities Nuclear Power Station, Units 1 and 2 (Reference 9), the Dresden Nuclear Power Station, Units 2 and 3 (Reference lo), and the Prairie Island Nuclear Generating Plant, Unit 2 (Reference 11).

Page 6 of 7

ENCLOSURE I 10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY REFERENCES

1. Regulatory Guide 1.147, "lnservice lnspection Code Case Acceptability, ASME Section XI, Division 1," Revision 14, August 2005.

2. ASME Section XI Code Case N-460, "Alternative Examination Coverage for Class 1 and Class 2 Welds."

3. ASME Section XI Code Case N-613-1, "Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-Dl Item No's. B3.10 and 83.90, Reactor Nozzle-To-VesselWelds, ~ i ~ u r e s 1 ~ ~ - 2 5 0 0 -(b),

7 ( aand

),

(c)."

4. NRC Information Notice 9842, "Implementation of 10 CFR 50.55a(g) In-service lnspection Requirements."

5. EPRl Internal Report IR-2004-63, "Monticello Nozzle lnner Radius and Nozzle-to-Shell Weld Examinations," dated December 2004.

6. EPRl Internal Report IR-2006-100, "Monticello Nozzle lnner Corner Regions and Nozzle-to-Shell Weld Examinations," dated January 2006.

7. BWRVIP-130, "BWR Water Chemistry Guidelines - 2004 Revision1'(EPRI Topical Report TR-1008192).

8. NRC letter to NMC, "Monticello Nuclear Generating Plant (MNGP) - Fourth 10-Year lnterval lnservice lnspection (ISI) Program Plan Relief Request No. 13 (TAC No. MC8882)," dated July 18, 2006.

9. Letter from NRC to Exelon Generation Company, LLC, "Quad Cities, Units 1 and 2 - Relief Request CR-39 for Third 10-Year lnservice lnspection lnterval (TAC Nos. MC2427 and MC2428)," dated May 10,2005.

10. Letter from NRC to Exelon Generation Company, LLC, "Dresden Nuclear Power Station, Units 2 and 3 - Relief Request CR-26 For Third 10-Year lnservice lnspection lnterval (TAC Nos. MC3269 and MC3270)," dated October 1, 2004.

11. NRC letter to NMC, "Prairie Island Nuclear Generating Plant, Unit 2 - Evaluation of Relief Request No. 16 for the Unit 2 3rd10-year lnterval lnservice lnspection Program (TAC No. MC1775)," dated October 18, 2004.

Page 7 of 7

ENCLOSURE 2 10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY TABLE A - Category B-D, "Full Penetration Welds of Nozzles in Vessels," ltem No. 83.90 Percent Coverage and Limitations for Nozzles N-2B, N9G, N-4A, N-6A, and N-9 Code System Code Component Category and Component and percent* Exam and Component 1D Examination Volume Coverage Report Item No. Description Required 0 btained Limitations Number Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D Recirculation Inlet N-2B NV Code Case N-613-1 78% 2007UT058 configuration.

B3.90 Nozzle N-2B Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle 6-D Recirculation lnlet N-2G NV Code Case N-6 13-1 78% 2007UT061 configuration.

B3.90 Nozzle N-2G Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D Feedwater Inlet N-4A NV Code Case N-613-1 79% 2007UT103 configuration.

83.90 Nozzle N-4A Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D Top Head Spare N-6A NV Code Case N-613-1 8696 2007UT104 configuration.

B3.90 Nozzle N-6A Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D CRD Return Nozzle N-9 NV Code Case N-613-1 85% 2007UTl02 configuration.

B3.90 (capped) N-9 Figure 2

• Due to the nozzle design it was not feasible to effectively examine essentially 100 percent of the required examination volume as defined in Figure 2 of Code Case N-613-1. Percentages are conservatively rounded down to the nearest whole number. It should be noted that I 0 0 percent of the inner 15 percent was examined for all components listed above.

Page 1 of 1

ENCLOSURE 3 10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY EXAM LIMITATIONS IMPOSED BY COMPONENT DESIGN AND CONSTRUCTION This enclosure contains a series of excerpts from the IS1 Ultrasonic Testing (UT) reports applicable to the subject components.

These excerpts contain sketches depicting the component configuration with physical limitations imposed by the design, e.g., geometrical contour, weld position, interferences, and a cross sectional view depicting the UT coverage and limitations in relation to the required examination volume.

Also included is a sketch of a typical reactor vessel nozzle contour and the resulting effect that causes the UT transducer to lift and lose effective coupling when it reaches the nozzle blend radius.

COMPONENT REPORT PAGE(S)

Pages 1-2 Pages 3-4 Page 5 Page 6 Page 7 Typical Reactor Vessel Nozzle Contour Affecting Page 8 Transducer Contact at blend radius 8 Pages Follow

Coverage drawings excerpted from applicable reports Component N-2B NV Report # 2007UT058 Supplemental Report Report No.: 2007UT058 Summary No.: 102658

-JP'\ Monticello N2 Coverage Plot Axial scan direction Page 1 of 8

Component N-2B NV Report # 2007UT058 Supplemental Report Report No.: 2007UT058 Summary No.: 102658 Monticello N2 Coverage Plot Parallel scan direction Inner 1 5%

Page 2 of 8

Component N-2G NV Report # 2007UT061 Supplemental Report Report No.: 2007UT061 Summary No.: 102668

- Monticello N2 Coverage Plot Axial scan direction Page 3 of 8

Component N-2G NV Report # 2007UT061 Supplemental Report Report No.: 2007UT061 Summary No.: 102668 Monticello N2 Coverage Plot Parallel scan direction Inner 1 5%

Page 4 of 8

Component N-4A NV Report # 2007UT103

- ;upplemental Report Report No.: 2007UT103 Summary No.: 102684 Comments: Coverage Plots R3.00in Monticello N4 Coverage Plot Axial scan direction N4 Coverage Plot ParalIel scan direction Page 5 of 8

Component N-6A NV Report # 2007UT104 Supplemental Report Report No.: 2007UT104 Summarv No.: 162375 Comments: N-6A NV Coverage Plots Monticello N6 Coverage Plot \ Monticello N6 Coverage Plot Axial scan direction \

Circ scan direction

<\,

"\

Inner 15%

'H k' F' 'ii Page 6 of 8

Component N-9 NV Report # 2007UT102 Supplemental Report Report No.: 2007UT102 Summary No.: 102700 Comments: Coverage Plots Monticello N9 Coverage Plot Monticello N9 Coverage Plot Axial scan direction Circ scan direction Page 7 of 8

Typical Representation of Nozzle Limitations Coverage affected by liftoff due to radius I\ Exit point o f 1 I transducer I Axial scan shown N2 Nozzle shown as example Page 8 of 8