GO2-14-017, Inservice Inspection (ISI) Program Request 3 ISI-14

Inservice Inspection (ISI) Program Request 3 ISI-14

ML14042A160
Person / Time
Site:	Columbia
Issue date:	02/06/2014
From:	Javorik A Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GO2-14-017
Download: ML14042A160 (7)

Text

X Alex Javorik ENERGY Vice P.O. BoxPresident, 968, MailEngineering Drop PE20 J NORTHW EST Richland, WA 99352-0968 Ph. 509-377-8555 F. 509-377-4317 aljavorik@energy-northwest.com February 6, 2014 G02-14-017 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397; INSERVICE INSPECTION (ISI) PROGRAM REQUEST 3 IS1-14

References:

(1) Letter dated April 19, 2013, Sher Bahadur (NRC) to Dennis Madison (BWRVIP), "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP-241) Report, 'Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-To-Vessel Shell Welds and Nozzle Blend Radii" (2) Letter dated December 19, 2007, Matthew A. Mitchell (NRC), to Rick Libra (BWRVIP), "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"'

Dear Sir or Madam:

Section 50.55a of Title 10 of the Code of Federal Regulations requires that In Service Inspection (ISI) of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 piping be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. Pursuant to 10 CFR 50.55a(a)(3)(i) Energy Northwest hereby requests NRC approval of the proposed alternate to ASME Section XI, Sub Article IWB-2500 to allow reduced percentage requirements for nozzle to vessel weld and inner radius examinations while still providing an acceptable level of quality and safety. This alternative is requested for the third ten-year interval ISI program at Columbia Generating Station. The details of the 10 CFR 50.55a request are enclosed as Attachment 1.

Approval of request 31S1-14 would allow reduced examination requirements through application of American Society of Mechanical Engineers (ASME) Code Case N-702. The NRC provided a Safety Evaluation approving the generic technical bases and acceptability criteria for application of Code Case N-702, which Energy Northwest has followed as detailed in the attached request.

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 Page 2 Energy Northwest requests approval by February 28, 2015 to accommodate application of the request during the next refueling outage. If approved, the use of Code Case N-702 at Columbia Generating Station would result in significantly reduced radiological dose to personnel while providing acceptable level of quality and safety.

There are no new commitments made in this submittal. If you have any questions or require additional information, please contact Lisa Williams at 509-377-8148.

Respectfully, A. L. Javorik Vice President, Engineering

Attachment:

(1) 10 CFR 50.55a Request Number 31S1-14 cc: NRC RIV Regional Administrator CF Lyon, NRC NRR Project Manager NRC Sr. Resident Inspector - 988C AJ Rapacz - BPA/1 399 WA Horin - Winston & Strawn

INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 1 of 5 10 CFR 50.55a Request Number 31S1-1 4 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

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

1. ASME Code Component(s) Affected The components in the following table are affected by this request.

Table 1 Identification Code Item Number Description Category Number N1-0 RRC NOZZLE TO VESSEL WELD @ 0 Deg B-D B3.90 N1 IR RRC NOZZLE INNER RADIUS @ 0 Deg B-D B3.1 00 N1 -180 RRC NOZZLE TO VESSEL WELD @ 180 Deg B-D B3.90 N1 -180-IR RRC NOZZLE INNER RADIUS @ 180 Deg B-D B3.100 N2-30 RRC NOZZLE TO VESSEL WELD @ 30 Deg B-D B3.90 N2-30-IR RRC NOZZLE INNER RADIUS @ 30 Deg B-D B3.100 N2-60 RRC NOZZLE TO VESSEL WELD @ 60 Deg B-D B3.90 N2-60-IR RRC NOZZLE INNER RADIUS @ 60 Deg B-D B3.100 N2-90 RRC NOZZLE TO VESSEL WELD @ 90 Deg B-D B3.90 N2-90-IR RRC NOZZLE INNER RADIUS @ 90 Deg B-D B3.100 N2-120 RRC NOZZLE TO VESSEL WELD @ 120 Deg B-D B3.90 N2-120-IR RRC NOZZLE INNER RADIUS @ 120 Deg B-D B3.100 N2-150 RRC NOZZLE TO VESSEL WELD @ 150 Deg B-D B3.90 N2-150-IR RRC NOZZLE INNER RADIUS @ 150 Deg B-D B3.100 N2-210 RRC NOZZLE TO VESSEL WELD @ 210 Deg B-D B3.90 N2-210-IR RRC NOZZLE INNER RADIUS @ 210 Deg B-D B3.100 N2-240 RRC NOZZLE TO VESSEL WELD @ 240 Deg B-D B3.90 N2-240-IR RRC NOZZLE INNER RADIUS @ 240 Deg B-D B3.100 N2-270 RRC NOZZLE TO VESSEL WELD @ 270 Deg B-D B3.90 N2-270-IR RRC NOZZLE INNER RADIUS @ 270 Deg B-D B3.100 N2-300 RRC NOZZLE TO VESSEL WELD @ 300 Deg B-D B3.90 N2-300-IR RRC NOZZLE INNER RADIUS @ 300 Deg B-D B3.100 N2-330 RRC NOZZLE TO VESSEL WELD @ 330 Deg B-D B3.90 N2-330-IR RRC NOZZLE INNER RADIUS @ 330 Deg B-D B3.100 RRC Reactor Recirculation

2. Applicable Code Edition and Addenda

Columbia Generating Station (Columbia) ASME Section Xl Code is the 2001 Edition through the 2003 Addenda (Reference 3). Additionally, for ultrasonic examinations, Section Xl, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is implemented as required (and modified) by 10 CFR 50.55a(b)(2)(xv).

INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 2 of 5

3. Applicable Code Requirement

Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are given in Subsection IWB, Table IWB-2500-1, "Examination Category B-D Full Penetration Welds of Nozzles in Vessels - Inspection Program B," Item Numbers B3.90 and B3.100, respectively. The method of examination is volumetric. For the extent of examination, all nozzles with full penetration weld to the vessel shell (or head) and integrally cast nozzles must be examined each interval.

4. Reason for Request

The identified nozzles (see Table 1) are scheduled for examination prior to the end of the current inspection interval for Columbia. The interval began in December 2005 and will end in December 2015. 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 1250 mRem for the next refuel outage.

5. 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 in Table 1 above. As an alternative, for all welds and inner radii identified in Table 1, Columbia proposes to examine a minimum of 25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each type and nominal pipe size, in accordance with Code Case N-702 (Reference 4). The proposed schedule to complete the minimum of 25% from each group is presented in Table 2 below. Columbia has completed three assembly (nozzle and inner radii) examinations already this interval using automated ultrasonic testing techniques. Greater than 97% coverage was achieved on each location and no relevant indications were reported. With the acceptance of the relief request only one nozzle-to-vessel weld and one inner radius would be required to meet the 25% for the current interval.

Table 2 Total Number to Group Number be Examined Comments RRC Outlet (N1) 2 1 1 Completed in R19 RRC Inlet (N2) 10 3 2 Completed in R19, 1 scheduled in R22 R19 Refuel Outage 19 (2009)

R22 Refuel Outage 22 (2015)

Basis for Use Boiling Water Reactor (BWR) Vessel Internals Project (BWRVIP) topical report BWRVIP-241, "BWR Vessel and Internals Project Probabilistic Fracture Mechanics Evaluation for the Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii" (hereafter referred to as BWRVIP-241)

(Reference 2), documents supplemental analyses for BWR reactor pressure vessel (RPV) recirculation inlet and outlet nozzle-to-shell welds and nozzle inner radii. BWRVIP-241 was submitted to address the limitations and conditions specified in the December 19, 2007, safety evaluation (SE) for the BWRVIP-1 08NP report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii" (Reference 1). The BWRVIP-1 08NP report contains the technical basis

INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 3 of 5 supporting American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," for reducing the inspection of RPV nozzle-to-vessel shell welds and nozzle inner radius areas from 100 percent to 25 percent of the nozzles for each nozzle type during each 10-year interval. Based on the two evaluations (BWRVIP-241 and BWRVIP-1 08NP), the failure probabilities due to a low temperature over pressure (LTOP) event at the nozzle blend radius region and the nozzle-to-vessel shell weld for Columbia recirculation nozzles are very low and meet the NRC safety goal.

Based on the results of this evaluation, the report concluded that the inspection of 25% of each nozzle type is technically justified as per Code Case N-702.

EPRI report BWRVIP-241 received a final NRC SE on April 19, 2013 (ML13071A240)

(Reference 6). In the SE, Section 5.0 "Conditions and Limitations" indicates that each licensee who plans to request relief from the ASME Code, Section Xl requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference BWRVIP-241 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 of the BWRVIP-241 report to their units in the relief request by demonstrating all of the following:

(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 115OF per hour.

The Columbia Technical Specification limits the heatup/cooldown rate to less than or equal to 100 OF in any one hour period.

For the Recirculation Inlet Nozzles (N2) the following criteria must be met:

(2) (pr/t)/CRPV*1l.15.

(3) [p(r02 +ri2)/(ro2-rI2)]/CNozzLE<1.47.

For the Recirculation Outlet Nozzles (N1) the following criteria must be met:

(4) (pr/t)/CRPV<1.15.

(5) [p (ro2+ri 2)/(r0 2-ri2 )]/CNozZLE<I .59.

The terms to be used in the NRC SE Section 5 applicability evaluations criteria 2-4 are:

CaPv = recirculation inlet nozzles N2 (from BWRVIP-241 model) = 19332 CNOZZLE = recirculation inlet nozzles N2 (from BWRVIP-241 model) = 1637 CRPV = recirculation outlet nozzles N1 (from BWRVIP-241 model) = 16171 CNOZZLE = recirculation outlet nozzles N1 (from BWRVIP-241 model) = 1977 p = RPV normal operating pressure (psi) r = RPV inner radius (inch) t = RPV wall thickness (inch) r= Nozzle inner radius (inch) ro = Nozzle outer radius (inch)

INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 4 of 5 Table 3 below summarizes these results.

Table 3 Outlet Nozzles (N1)

Criteria (4) Criteria (5)

Cnozzle CRPV p r t(min) ri ro <1.15 <1.59 1977 16171 1020 127 9.5 10.8 15.4 0.84 1.51 Inlet Nozzles (N2)

Criteria (2) Criteria (3)

Cnozzle CRPV p r t(min) ri r, <1.15 <51.47 1637 19332 1020 127 9.5 5.8 10.0 0.71 1.25 Based upon the above information, the RRC Inlet and Outlet nozzle-to-vessel shell welds and nozzle inner radii sections identified in Table 1 meet the SE criteria and therefore Code Case N-702 is applicable. In addition, the reactor vessel is low alloy steel plate specification SA-533 grade B class I. The nozzle is low alloy steel forging specification SA-508 class 2. The weld metal used in the welds specified in Table 1 is carbon/low alloy steel and the welds are outside the beltline fluence region. 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 RRC Inlet and Outlet nozzle-to-vessel shell welds and nozzle inner radii sections identified in Table 1.

6. Duration of Proposed Alternative The duration of this request is for the third inservice inspection interval ending December 12, 2015.

7. Precedents None

8. References

1. BWRVIP-IO8NP: 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 Blend Radii. EPRI, Palo Alto, CA: 2007.1016123.

2. BWRVIP-241: BWR Vessel and Internals Project, ProbabilisticFractureMechanics Evaluation for the Boiling Water Reactor Nozzle-to- Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2010. 1021005.

3. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plants," 2001 Edition through 2003 Addenda.

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

INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31S1-14 ATTACHMENT 1 Page 5 of 5

5. Matthew A. Mitchell, Office of Nuclear Reactor Regulation, to Rick Libra, BWRVIP Chairman, "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-1 08)',"

December 19, 2007.

6. Sher Bahadur, Office of Nuclear Reactor Regulation, to Dennis Madison, BWRVIP Chairman, "Final Safety Evaluations of the Boiling Water Reactor Vessel Internals Project (BWRVIP-241) Report, 'Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-To-Vessel Shell Welds and Nozzle Blend Radii (TAC NO.

ME6328)" April 19, 2013.