NLS2010014, Request Relief from Certain Inservice Inspection Code Requirements

Request Relief from Certain Inservice Inspection Code Requirements

ML100470703
Person / Time
Site:	Cooper
Issue date:	02/05/2010
From:	O'Grady B Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2010014
Download: ML100470703 (8)

Text

Nebraska Public Power District "Always there when you need us" 10OCFR 50.55a NLS2010014 February 5, 2010 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

10 CFR 50.55a Request Number RI-04, Revision 0 Cooper Nuclear Station, Docket No. 50-298, DPR-46

Dear Sir or Madam:

The purpose of this letter is to request that the Nuclear Regulatory Commission (NRC) grant Nebraska Public Power District (NPPD) relief from certain inservice inspection (ISI) code requirements for Cooper Nuclear Station (CNS) pursuant to 10 CFR 50.55a.

10 CFR 50.55a Request Number RI-04, Revision 0 is applicable to the fourth ten-year ISI interval, which began March 1, 2006. NPPD requests NRC approval of the attached request by February 8, 2011, which represents a, standard twelve-month review period following tlle submittal. Approval of this request is needed to support examinations scheduled for Refueling Outage 26.

RI-04, Revision 0 is contained in the attachment to this letter.

If you have any questions concerning this matter, please contact David Van Der Kamp, Licensing Manager, at (402) 825-2904.

Sii Brian J. O'Grady Site Vice President

/dm Attachment COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 04ý www.nppd.com

NLS2010014 Page 2 of 2 cc: Regional Administrator w/ attachment USNRC - Region IV Cooper Project Manager w/ attachment USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachment USNRC - CNS NPG Distribution w/ attachment CNS Records w/ attachment

NLS2010014 Attachment Page 1 of 5 10CFR 50.55a Request Number RI-04, Revision 0 Inspection of Reactor Vessel Nozzle-to-Vessel Shell Welds and Nozzle Inner-Radius Sections Cooper Nuclear Station Docket No. 50-298, DPR-46 Proposed Alternative in Accordance to 10CFR50.55a(a)(3)(i)

--Alternative Provides Acceptable Level of Quality and Safety--

COMPONENT IDENTIFICATION Code Class: 1 Examination Category: B-D Item Number: B3.90, B3.100

==

Description:==

Inspection of Reactor Vessel Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius Sections Component Numbers:

Table RI-04-1 Specific Components that Relief is Requested Nozzle-to-Vessel Shell Welds that Relief is Requested ASME Item No. Component ID Item Description B3.90 NVE-BD-N1B 28" Recirculation Outlet B3.90 NVE-BD-N5B 10" Core Spray B3.90 NVE-BD-N6A 6" Instrumentation B3.90 NVE-BD-N8A 5" Jet Pump Instrumentation B3.90 NVE-BD-N3B 24" Main Steam B3.90 NVE-BD-N3C 24" Main Steam B3.90 NVE-BD-N3D 24" Main Steam Nozzle Inner Radius Sections that Relief is Requested ASME Item No. Component ID Item Description B3.100 NVIR-BD-N1B 28" Recirculation Outlet B3.100 NVIR-BD-N5B 10" Core Spray B3. 100 NVIR-BD-N6A 6" Instrumentation B3.100 NVIR-BD-N8A 5" Jet Pump Instrumentation B3. 100 NVIR-BD-N3B 24" Main Steam B3. 100 NVIR-BD-N3C 24" Main Steam B3.100 NVIR-BD-N3D 24" Main Steam

NLS2010014 Attachment Page 2 of 5

Applicable Code Edition and Addenda

The applicable Code Edition and Addenda for Cooper Nuclear Station (CNS) is American Society of Mechanical Engineers (ASME) Code Section XI, 2001 Edition, 2003 Addenda.

Additionally for ultrasonic examinations,Section XI, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems," is implemented as required and subject to the limitations and modifications specified in 10CFR50.55a(b)(2)(xv).

Applicable Code Requirement

ASME Section XI Code Class 1 nozzle-to-vessel shell weld and nozzle inner radius section examination requirements are provided in Subsection IWB, Table IWB-2500-1 "Examination Category B-D, Full Penetration Welded Nozzles in Vessels - Inspection Program B." Items B3.90 and B3. 100 require a volumetric examination of all the Reactor Vessel nozzle-to-vessel welds and associated nozzle inner radius sections, respectively.

Reason for Request

The identified nozzles (see Table RI-04-1) are scheduled for examination prior to the end of the current inspection interval. The proposed alternative provides an acceptable level of quality and safety, and the reduction in scope could provide a dose savings of as much as 15.5 Radiation Equivalent Man over the remainder of the interval.

Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from performing the required examinations on 100% of the identified nozzle assemblies (see Table below). As an alternative, for all welds and nozzle inner radius sections identified in Table RI-04-1, CNS proposes to examine a minimum of 25% of the nozzle-to-vessel shell welds and nozzle inner radius sections to include at least one nozzle from each system and nominal pipe size, in accordance with Code Case N-702 (Reference 3). For the nozzle assemblies identified this would mean one from each of the groups identified below:

Table RI-04-2 Total Group/Component ID Number No. to be Examined Nozzles Recirculation Outlet (NI) 2 1 Recirculation Inlet (N2, does not pass criteria) 10 10 Main Steam (N3) 4 1 Feedwater (N4, not allowed by Code Case) 4 4 Core Spray (N5) 2 1 Head Instrument (N6) 2 1 Head Vent (N7) 1 1

NLS2010014 Attachment Page 3 of 5 Table RI-04-2 Total Group/Component ID Number No. to be Examined Nozzles Jet Pump (N8) 2 1 Control Rod Drive Return (N9, not allowed by Code Case)

Code Case N-702 stipulates that Visual Test (VT) VT-I examination may be used in lieu of the volumetric examination for the nozzle inner radius sections (Item No. B3.100). CNS currently credits the enhanced magnification VT-I examination of the nozzle inner radius sections in accordance to Code Case N-648-1 (Reference 4) subject to the conditions placed upon the use of that Code Case by Regulatory Guide 1.147. The specific aspect of utilizing VT-I visual examinations as allowed by Code Case N-702 is not part of this request. Volumetric examinations of the nozzle inner radius sections of the selected core spray and jet pump instrumentation nozzles will still be performed, as their nozzle inner radius sections are not fully accessible from inside the vessel for enhanced magnification VT-I examination.

Electric Power Research Institute (EPRI) Technical Report 1003557 (Reference 1), "BWRVIP-108: Boiling Water Reactor 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," provides the basis for Code Case N-702. The evaluation found that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell welds due to a Low Temperature Overpressure event are very low (i.e., <1 x 10-6 for 40 years) with or without inservice inspection. The report concludes that inspection of 25% of each nozzle type is technically justified.

The Nuclear Regulatory Commission issued a Safety Evaluation Report (SER) dated December 19, 2007 (Reference 2) approving the use of BWRVIP- 108 as a basis for using Code Case N-702. In the SER, Section 5.0 "Plant Specific Applicability" states that licensees who plan to request relief from the ASME Code,Section XI requirements for reactor pressure vessel (RPV) nozzle-to vessel shell welds and nozzle inner radius sections may reference the BWRVIP- 108 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability for the BWRVIP- 108 report to its units in the relief request by showing that all the general and nozzle specific criteria are satisfied:

Maximum RPV Heatup/Cooldown Rate (1) the maximum RPV heatup/cooldown rate is limited to less than 1 15'F per hour.

This criterion is met by adherence to CNS Technical Specifications Surveillance Requirement 3.4.9.1 which requires verification that the Reactor Coolant System heatup and cooldown rates are <I00°F when averaged over a one hour period.

NLS2010014 Attachment Page 4 of 5 For recirculation inlet N2 nozzles, Criteria (2) and (3) of the SER apply:

(2) (pr/t)/Ci-Rpv< 1.15 where:

p = RPV normal operating pressure (p= 1020 psig per CNS Technical Specifications 3.4.10 for Reactor Steam Dome Pressure) r = RPV inner radius (r = 110.375 in.)

t = RPV wall thickness (t = 6.875 in.)

Ci-RPV = 19332 (based on the BWRVIP-108 recirculation inlet nozzle/RPV finite element method (FEM) model)

The CNS result is 0.847, which is less than 1.15, therefore the CNS N2 nozzles meet criteria 2.

(3) [p(ro2 +ri 2) / (ro2-ri2)] / Ci-NOZZLE < 1.15 where:

p = RPV normal operating pressure (p = 1020 psig) r0 = nozzle outer radius (ro = 10.219 in.)

ri = nozzle inner radius (ri= 6.188 in.)

Ci-NOZZLE = 1637 based on the BWRVIP-108 recirculation inlet nozzle/RPV FEM model)

The calculation result is 1.344, which is greater than 1.15, therefore the CNS N2 nozzles do not meet Criteria 3.

For recirculation outlet NI nozzles, Criteria (4) and (5) of the SER apply:

(4) (pr/t)/Co.yv< 1.15 where:

p = RPV normal operating pressure (p = 1020 psig) r = RPV inner radius (r = 110.375 in.)

t = RPV wall thickness (t = 6.875 in.)

Co-RPV = 16171 (based on the BWRVIP- 108 recirculation outlet nozzle/RPV FEM model)

The calculation result is 1.013, which is less than 1.15, therefore the CNS NI nozzles meet Criteria 4.

(5) [p(r 0 2 +r, 2)/ (r 0 2 _-r2 )] / Co-NOZZLE < 1.15 where:

p = RPV normal operating pressure (p = 1020 psig) r0 = nozzle outer radius (ro = 21.656 in.)

ri = nozzle inner radius (ri= 12.875 in.)

Co-NOZZLE = 1977 (based on the BWRVIP- 108 recirculation outlet nozzle/RPV FEM model)

The calculation result is 1.080, which is less than 1.15 therefore the CNS NI nozzles meet Criteria 5.

Based upon the above information, all RPV nozzle-to-vessel shell welds and nozzle inner radius sections, with the exception of the Recirculation Inlet N2 Nozzles, meet the criteria. Therefore,

NLS2010014 Attachment Page 5 of 5 Code Case N-702 is applicable. Since the Recirculation inlet nozzles do not meet all of the criteria, Code Case N-702 will not be applied to these nozzles.

Therefore, use of Code Case N-702 provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(a)(3)(i) for all RPV nozzle-to-vessel shell welds and nozzle inner radius sections, with the exception of the Recirculation Inlet Nozzles.

Duration of Proposed Alternative The proposed alternative is for the fourth ten-year interval of the Inservice Inspection Program for CNS that started on March 1, 2006 and ends in 2014, at the end of the current license.

References

1. EPRI Technical Report 1003557, "BWRVIP-108: Boiling Water Reactor 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," October 2002.

2. Safety Evaluation of Proprietary EPRI Report, "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 Inner Radius (BWRVIP-108)," December 19, 2007.

3. ASME Section XI Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI, Division 1," February 20, 2004.

4. ASME Section XI Code Case N-648-1, "Alternative Requirements for Inner Radius Examination of Class I Reactor Vessel NozzlesSection XI, Division 1," September 7, 2001.

4 ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS©4 Correspondence Number: NLS2010014 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

'COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None N/A N/A t I f I I 4 PROCEDURE 0.42 REVISION 24 PAGE 18 OF 25