ML082950482
| ML082950482 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 10/09/2008 |
| From: | Gambhir S Energy Northwest |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| G02-08-142 | |
| Download: ML082950482 (9) | |
Text
ENERGY NORTHWEST Sudesh K. Gambhir Vice President, Technical Services P.O. Box 968, Mail Drop PE04 Richland, WA 99352-0968 Ph. 509-377-8313 F. 509-377-2354 sgambhir@energy-northwest.com October 9, 2008 G02-08-142 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001
Subject:
Reference:
COLUMBIA GENERATING STATION, DOCKET NO. 50-397 INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31SI-09 Letter dated December 19, 2007, MA Mitchell (NRR) to R Libra (BWRVIP),
"Safety Evaluation of Proprietary 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)"'
Dear Sir or Madam:
Section 50.55a of Title 10 of the Codeof Federal Regulations requires -that Inservice Inspection (ISI) of American Society of Mechanical -Engineers (ASME) Code Class 1, 2, and 3 piping be performed in accordance withý SectiOn.Xi of the ASMEý Boiler and' Pressure Vessel Code. Pursuant to 10 CFR 50.55a(a)(3)(i) Energy Northwest hereby requests NRC approval of the alternative (Attachment 1) to ASME Section XI, Sub Article IWB-2500 to allow reduced percentage requirements for nozzle to vessel weld and inner radius examinations. This alternative is requested for the third ten-year interval ISI program at Columbia Generating Station.
Approval of request 31SF-09 would allow reduced examination requirements through application 6f American Society of Mechanica[-Engineers-(ASME) Code Case N702.
The NRC: has provided a Safety Evaluation approving the generic technical basis and acceptability criteria for application of Code Case 702, which Energy Northwest has followed as detailed in the attached request.
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31SI-09 Page 2 Energy Northwest requests approval by May 1, 2009 to accommodate application of the request during the next refueling outage. If approved, the use of Code Case 702 at Columbia Generating Station would result in significantly reduced radiological dose to personnel while providing an 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 MC Humphreys at 509 377-4025.
Rpectfully, Gambhir Vice President, Technical Services Attachments:
(1) 10 CFR 50.55a Request Number 31SI-09 (2) Plant Specific Applicability cc: EE Collins, Jr. - NRC RIV CF Lyon - NRC NRR NRC Senior Resident Inspector/988C WA Horin - Winston & Strawn RN Sherman - BPA/1 399
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31SI-09 Page 1 of 6 10 CFR 50.55a Request Number 31SI-09 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/Azimuth Category Number N3-72 MS Nozzle to Vessel Weld @ 72 B-D B3.90 N3-72-IR MS Nozzle Inner Radius @ 72 B-D B3.100 N3-108 MS Nozzle to Vessel Weld @ 108 B-D B3.90 N3-108-IR MS Nozzle Inner Radius @ 108 B-D B3.100 N3-252 MS Nozzle to Vessel Weld @ 252 B-D B3.90 N3-252-IR MS Nozzle Inner Radius @ 252 B-D B3.100 N3-288 MS Nozzle to Vessel Weld @ 288 B-D B3.90 N3-288-IR MS Nozzle Inner Radius @ 288 B-D B3.100 N5-120 LPCS Nozzle to Vessel Weld @ 120 B-D B3.90 N5-120-IR LPCS Nozzle Inner Radius @120 B-D B3.100 N6-45 LPCI Nozzle to Vessel Weld @ 45 B-D B3.90 N6-45-IR LPCI Nozzle Inner Radius @ 45 B-D B3.100 N6-135 LPCI Nozzle to Vessel Weld @ 135 B-D B3.90 N6-135-IR LPCI Nozzle Inner Radius @135 B-D B3.100 N6-315 LPCI Nozzle to Vessel Weld @ 315 B-D B3.90 N6-315-IR LPCI Nozzle Inner Radius @ 315 B-D B3.100 N9-105 JP Instrumentation Nozzle to Vessel B-D B3.90 Weld @ 105 N9-105-IR JP Instrumentation Nozzle Inner B-D B3.100 Radius@ 105 N9-285 JP Instrumentation Nozzle to Vessel B-D B3.90 Weld @ 285 N9-285-IR JP Instrumentation Nozzle Inner B-D B3.100 Radius@ 285 N16-240 HPCS Nozzle to Vessel Weld @ 240 B-D B3.90 N16-240-1R HPCS Nozzle Inner Radius @ 240 B-D B3.100 N7 Top Head Spray Nozzle to Top B-D B3.90 Head Weld
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31SI-09 Page 2 of 6 Table 1 Identification Code Item Number Description/Azimuth Category Number N7-IR Top Head Spray Nozzle Inner B-D B3.100 Radius N8 Top Head Vent Nozzle to Top Head B-D B3.90 Weld N8-IR Top Head Vent Nozzle Inner Radius B-D B3.100 N18 Top Head Spare Nozzle to Top B-D B3.90 Head Weld N18-IR Top Head Spare Nozzle Inner B-D B3.100 Radius JP Jet Pump LPCS Low Pressure Core Spray LPCI Low Pressure Core Injection HPCS High Pressure Core Spray MS Main Steam
2. Applicable Code Edition and Addenda
The applicable Code Edition and Addenda for Columbia Generating Station (Columbia) is ASME Section Xl Code 2001 Edition through the 2003 Addenda (Reference 2).
Additionally, for ultrasonic examinations,Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," is implemented as required (and modified) by 10 CFR 50.55a(b)(2)(xv).
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 Welded Nozzles in Vessels - Inspection Program B," Item Numbers B3.90 and B3.100 "Nozzle Inside Radius Section," respectively. The method of examination is volumetric. With respect to the extent of examination, all nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles must be examined each interval. All of the nozzle assemblies identified in Table 1 are full penetration welds.
4. Reason for Request
The proposed alternative provides a reduction in refuel outage work scope and could provide a dose savings of as much as 2.0 Rem for the next refuel outage and 3.0 Rem over the remainder of the interval. The identified nozzles (see Table 1) are scheduled for examination prior to the end of the current inspection interval for Columbia.
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31SI-09 Page 3 of 6
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, Energy Northwest proposes to examine a minimum of 25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle/inner radius section from each system (except recirculation) and nominal pipe size, in accordance with Code Case N-702 (Reference 3). For the components identified in Table 1, this would mean at least one nozzle/inner radius section from each of the groups identified in Table 2 will be examined.
Table 2 Total Number to be Group Number Examined Comments Main Steam (N3) 4 1
1 scheduled in R-20 Core Spray (N5, N16) 2 1
1 scheduled in R-22 Reactor Low Pressure 3
1 1 scheduled in R-22 Injection (LPCI) (N6)
Jet Pump (N9) 2 1
1 scheduled in R-22 Top Head Nozzles 3
1 1 scheduled in R-20 (N7, N8, N18)
R-20 Refuel Outage 20 (2011).
R-22 Refuel Outage 22 (2015).
Nominal pipe size is uniform within each group.
Code Case N-702 stipulates that VT-1 examination may be used in lieu of the volumetric examination for the inner radii (Item No. B3.1 00). Note that Energy Northwest is not currently using Code Case N-648-1 on enhanced magnification visual examination and has no plans of using Code Case N-648-1 in the future. All examinations on Item B3.100 will be volumetric examinations.
Electric Power Research Institute (EPRI) Technical Report 1003557, "BWRVIP-1 08:
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" (References 1 and 4), provides the basis for Code Case N-702. From the evaluation, the report concluded that the failure probabilities due to a Low Temperature Over-Pressurization event at the nozzle blend radius region and the nozzle-to-vessel shell weld 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.
This EPRI report was approved by an NRC Safety Evaluation (SE) dated December 19, 2007 (Reference 5). In Section 5.0 "Plant Specific Applicability," the SE indicates-that each licensee who plans to request relief from the ASME Code,Section XI
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31SI-09 Page 4 of 6 requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-1 08 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-1 08 report to its units in the relief request by showing that all the general and nozzle specific criteria addressed below are satisfied.
For all nozzles and inner radii the following general criterion (1) must be met:
(1)
The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 1150 F per hour.
This criterion is met by adherence to Columbia's Technical Specifications Surveillance Requirement 3.4.11.1 which requires verification that the Reactor Coolant System heatup and cooldown rate is less than or equal to 1000 F in any one hour period.
Nozzle-specific criteria (2) and (3) apply only to the Recirculation Inlet Nozzles (N2):
(2)
(pr/t)/CN2-RPV <1.15 The calculation for Columbia's N2 Nozzles results in 0.70 which is less than 1.15 and meets criteria 2.
(3)
[p(ro2 + ri2)/(ro2 - ri2)]/CN2.NOZZLE <1.15 The calculation for Columbia's N2 Nozzles results in 1.32 which is greater than 1.15 and does not meet criteria 3.
Nozzle-specific criteria (4) and (5) apply only to the Recirculation Outlet Nozzles (N1):
(4)
(pr/t)/CN1-RPV < 1.15 The calculation for the Columbia's N1 Nozzles results in 0.84 which is less than 1.15 and meets criteria 4.
(5)
[p(ro2 + ri2)/(ro2 - ri2)]/CN1_NOZZLE <1.15 The calculation for Columbia's N1 Nozzles results in 1.61 which is greater than 1.15 and does not meet criteria 5.
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31SI-09 Page 5 of 6 The terms used in (2) through (5) above are defined as:
CN2-RPV "- recirculation inlet nozzles N2 constant (from BWRVIP-108 model) =
19332 psi CN2-NOZZLE = recirculation inlet nozzles N2 constant (from BWRVIP-108 model) =
1637 psi CN1-RPV = recirculation outlet nozzles N1 constant (from BWRVIP-108 model) =
16171 psi CN1-NOZZLE = recirculation outlet nozzles N1 constant (from BWRVIP-108 model) =
1977 psi r = RPV inner radius t = RPV wall thickness p = RPV normal operating pressure r = Recirculation nozzles inner radius
- r. = Recirculation nozzles outer radius Based upon application of the above criteria, all RPV nozzle-to-vessel shell welds and nozzle inner radii sections identified in Table 1 meet the criteria. Therefore Code Case N-702 is applicable. However, the Recirculation Inlet (N2) and Outlet (N1) nozzles do not meet all of the criteria and therefore Code Case N-702 would not be applied to recirculation system nozzles. The assumption, in section 5.0 of the SE transmitted by Reference 5, that only recirculation inlet and outlet nozzles need to be checked because the conditional probability of failure (P(FIE)) for other nozzles is an order of magnitude lower, remains valid. See Attachment 2 for details.
Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), use of Code Case N-702 provides an acceptable level of quality and safety for all RPV nozzle-to-vessel shell welds and nozzle inner radii sections, with the exception of Recirculation Inlet (N2) and Outlet (N1) nozzles. Table 1 identifies the population of welds to which the proposed relief would be applied.
6. Duration of Proposed Alternative
The duration of this request is for the third inservice inspection interval ending December 12, 2015. The use of Code case N-702 is requested-until the NRC publishes the Code case in a future revision of the applicable Regulatory Guide.
7. Precedents
A precedent for this request exists in the staffs SE approving this alternative for use by Duane Arnold Energy Center that was transmitted by Reference 6.
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31SI-09 Page 6 of 6
- 8. References
- 1. EPRI Technical Report 1003557, "BWRVIP-108: 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,"
October 2002.
- 2. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plants," 2001 Edition through 2003 Addenda.
- 3. 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.
- 4. BWRVIP letter 2002-323, Carl Terry, BWRVIP Chairman, to NRC Document Control Desk, "Project No. 704-BWRVIP-108: 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,"
November 25, 2002.
- 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. Letter, Lois James (NRR) to Richard L. Anderson (Duane Arnold Energy Center),
"Duane Arnold Energy Center - Safety Evaluation for Request for Alternative to Reactor Pressure Vessel Nozzle to Vessel Weld and Inner Radius Examinations (TAC NO. MD8193)," dated August 29, 2008.
INSERVICE INSPECTION (ISI) PROGRAM REQUEST 31SI-09 Page 1 of 1 Plant Specific Applicability Criteria 1 - The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 1150 F per hour.
Technical Specifications Surveillance Requirement 3.4.11.1 requires verification that the Reactor Coolant System heatup and cooldown rates are less than or equal to 100' F in any one hour period.
Criteria 2 - N2 Inlet criteria (pr/t)/ CRPV< 1.15 p = normal operating pressure 1020 psi r = RPV inner radius 126.7 in.
t =RPV wall thickness 9.5 in.
CRPV 19332 (pr/t)/ CRPV 0.70 Meets N2 Inlet criteria Criteria 3 - N2 Inlet Criteria [p(ro2 + ri2)/(ro2 - ri2)]/CN2.NOzZLE < 1.15 p = normal operating pressure 1020 psi ro = nozzle outer radius 10 in.
ri = nozzle inner radius 6 in.
CNOZZLE 1637
[p(ro2 +ri 2)/(ro 2-ri 2 )]/CNOzZLE 1.32 Does not meet N2 Inlet criteria Criteria 4 - N1 Outlet Criteria (pr/t)/ CRPV < 1.15 p = normal operating pressure 1020 psi r = RPV inner radius 126.7 in.
t = RPV wall thickness 9.5 in.
CRPV 16171 (pr/t)/ CRPV 0.84 Meets N1 Outlet criteria Criteria 5 - N1 Outlet Criteria [p(ro2 + ri2)/(ro2 - ri2)]/CN1-NOZZLE < 1.15 p = normal operating pressure 1020 psi ro = nozzle outer radius 15.36 in.
ri = nozzle inner radius 11.03 in.
CNOZZLE 1977
[p(ro2 + ri2)/(ro2 - ri2 )]/CNOzZLE 1.61 Does not meet N1 Outlet criteria