ML12122A980
| ML12122A980 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/22/2012 |
| From: | Istvan Frankl Plant Licensing Branch III |
| To: | O'Connor T Northern States Power Co |
| Tam P | |
| References | |
| TAC ME8087, RR-002 | |
| Download: ML12122A980 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 22, 2012 Mr. Timothy J. O'Connor Site Vice President Monticello Nuclear Generating Plant Northern States Power Company - Minnesota (NSPM) 2807 West County Road 75 Monticello, MN 55362-9637
SUBJECT:
MONTICELLO NUCLEAR GENERATION PLANT (MNGP) - RELIEF REQUEST RR-002, FOR THE FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL (TAC ME8087)
Dear Mr. O'Connor:
By letter dated September 28, 2011, Northern States Power Company - Minnesota (the licensee) requested changes to the inservice inspection program for the fifth 10-year inspection interval for Monticello Nuclear Generating Plant. Relief Request RR-002 in this submittal would revise the inspection requirements for certain reactor pressure vessel nozzle-to-vessel welds and nozzle inner radii sections from those based on American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, to an alternative based on ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor Nozzle Inner Radius and Nozzle-to-Shell Welds," without using the Code Case-specified visual (VT -1) examination.
The Nuclear Regulatory Commission (NRC) staff has completed review of NSPM's submittal; details of the review can be found in the enclosed safety evaluation. The NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i), and the licensee is in compliance with the ASME Code's requirements. Therefore, the NRC authorizes the licensee's proposed alternative for inspection of nozzle-to-vessel shell welds and nozzle inner radii sections of reactor pressure vessel nozzles speCified in Relief Request RR-002 for MNGP's fifth 10-year lSI interval.
T. J. O'Connor
- 2 Should you have any questions, please contact Mr. Peter Tam, the MNGP Project Manager, at 301-415-1451.
Docket No. 50-263
Enclosure:
Safety Evaluation cc w/encl: Distribution via ListServ Sincerely,
- li=-- ()Z&;J Istvan Frankl, Acting Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST RR-002 FOR RENEWED FACILITY OPERATING LICENSE NO. DPR-22 MONTICELLO NUCLEAR GENERATING PLANT NORTHERN STATES POWER COMPANY - MINNESOTA DOCKET NO. 50-263
1.0 INTRODUCTION
By letter dated September 28,2011 (Accession No. ML112720147), Northern States Power Company - Minnesota (the licensee) requested changes to the inservice inspection (lSI) program for the fifth 10-year inspection interval for Monticello Nuclear Generating Plant (MNGP). The Nuclear Regulatory Commission (NRC) staff has completed its review of Relief Request RR-002.
Relief Request RR-002 in the September 28, 2011, submittal would revise the inspection requirements for certain reactor pressure vessel (RPV) nozzle-to-vessel welds and nozzle inner radii sections from those based on American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, to an alternative based on ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," without using the Code Case-specified visual (Vf-1) examination.
2.0 REGULATORY EVALUATION
lSI of the ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as a way to detect anomaly and degradation indications so that structural integrity of these components can be maintained. This is required by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g), except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i). The provision of 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if: (i) the proposed alternatives would provide an acceptable level of quality and safety, or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
For all RPV nozzle-to-vessel shell welds and nozzle inner radii, ASME Code,Section XI requires 100 percent inspection during each 10-year lSI interval. However, ASME Code Case N-702 proposes an alternative which reduces 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 1 O-year interval. The NRC staff has previously approved the topical Enclosure
- 2 report BWRVIP-108, "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 Inner Radii," which contains the technical basis supporting ASME Code Case N-702. The NRC staff's December 19, 2007 (Accession No. ML073600374), safety evaluation (SE) regarding BWRVIP-108 specified plant-specific requirements which must be satisfied by licensees who propose to use ASME Code Case N-702.
The NRC staff has approved similar applications for several plants, including Dresden Nuclear Power Station, Units 2 and 3, Hope Creek Generating Station, and Cooper Nuclear Station.
3.0 TECHNICAL EVALUATION
The NRC staff's December 19, 2007, SE specified plant-specific requirements which must be met for licensees proposing to use this alternative. The licensee's September 28, 2011, submittal intends to demonstrate that the relevant MNGP RPV nozzle-to-vessel welds and the inner radii meet these plant-specific requirements so that Relief Request RR-002 can be approved.
As stated in the NRC staffs December 19, 2007, SE, licensees can demonstrate the plant specific applicability of the BWRVIP-108 report to their units in relief requests by meeting the following general and nozzle-specific criteria:
(1) The maximum RPV heatup/cooldown rate is limited to less than 115 of/hour; For recirculation inlet nozzles (2) (pr/t)/CRPv < 1.15 p = RPV normal operating pressure (psi),
r = RPV inner radius (inch),
t =RPV wall thickness (inch), and CRPV =19332; (3) [p(ro2+ r?)/ (ro2-r?)]/CNOZZLE< 1.15 p = RPV normal operating pressure (psi),
ro = nozzle outer radius (inch),
rj =nozzle inner radius (inch), and CNOZZLE = 1637; For recirculation outlet nozzles (4) (pr/t)/CRPV < 1.15 p =RPV normal operating pressure (psi),
r = RPV inner radius (inch),
t = RPV wall thickness (inch), and CRPV= 16171; and
- 3 p = RPV normal operating pressure (psi),
ro = nozzle outer radius (inch),
rj =nozzle inner radius (inch), and CNOZZLE =1977.
This plant-specific information was required by the NRC staff to ensure that the probabilistic fracture mechanics (PFM) analysis documented in the 8WRVIP-108 report applies to the RPV of the applicant's plant.
3.1 Licensee's Proposed Alternative The license requested the alternative for the following ASME Code Class 1 components:
Reactor Recirculation Inlet Nozzles - N2A, N28, N2C, N20, N2E, N2F, N2G, N2H, N2J, and N2K Main Steam Nozzles - N3A, N38, N3C, and N30 Core Spray Nozzles - NSA and NS8 RPV Head Nozzles - N6A and N68 Jet Pump Instrumentation Nozzles - N8A and N88 Note that the feedwater nozzles, and control rod drive return nozzles were not included in the licensee's request.
The affected examination category is 8-0, "Full Penetration Welded Nozzles in Vessels," and the examination item numbers are 83.90, "Nozzle-to-Vessel Welds" and 83.100, "Nozzle Inside Radius Section." The applicable requirement is at the 2007 Edition with the 2008 Addenda (i.e.,
the applicable lSI Code of Record for the fifth 10-year lSI interval for Monticello) of ASME Code,Section XI, Table IW8-2S00-1, "Examination Category 8-0, Full Penetration Welded Nozzles in Vessels." The licensee stated that:
Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are given in Item Numbers 83.90 "Nozzle-to-Vessel Welds" and 83.100 "Nozzle Inside Radius Section." The required method of examination is volumetric. All nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles require examination each interval.
The licensee proposed that:
As an alternative for examination of welds and inner radii on all the nozzle assemblies identified in Table S-1, Northern States Power - Minnesota (NSPM) proposes to examine a minimum of 2S% of the MNGP nozzle-to-vessel welds and inside radius sections, including at least one nozzle from each system and nominal pipe size, in accordance with ASME Code Case N-702. For the nozzle
- 4 assemblies identified in Table 5-1, this would mean that examination would be required for three of the Recirculation Inlet (N2) nozzles and one from each of the other nozzle groups, as identified below.
ASME Code Case N-702 stipulates that a VT-1 examination may be used in lieu of the volumetric examination for the inner radii (i.e., Item No. B3.100, "Nozzle Inside Radius Section"). However, NSPM is only requesting to perform volumetric examination of the applicable nozzle inner radius sections. NSPM is not requesting use of the VT-1 examination provisions included in the code case in lieu of performing volumetric examinations.
The licensee stated the bases for this alternate as follows:
[The BWRVIP-108 report] provides the basis for ASME Code Case N-702. The evaluation found that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld 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.
This EPRI report was approved by the NRC in a safety evaluation (SE) dated December 19, 2007 (Reference 3). Section 5.0, "Plant Specific Applicability," of the SE indicates that each licensee who plans to request relief from ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRIP-108 report as the technical basis for the use of ASME Code Case N-702 as an alternative. The NRC SE further states that each licensee should demonstrate the plant specific applicability criteria from the BWRVIP-108 report to its units in the relief request by showing that all the following general and nozzle-specific criteria are satisfied (i.e., as described in Attachment 2).
The licensee then concluded that:
Based on the information contained in Attachment 2, the nozzles listed in Table 5-1 have been demonstrated to meet the general and nozzle-specific criteria in BWRVIP-108 and the NRC SE as described above. As calculated in, the recirculation outlet or suction nozzles (N-1) do not meet the specified criteria and are excluded from this alternative.
The licensee proposed that the alternative be used for the entire fifth interval of the MNGP lSI Program scheduled to begin on September 1, 2012.
3.2
NRC Staff Evaluation
The December 19, 2007, SE for the BWRVIP-108 report specified five plant-specific criteria that a licensee must meet to demonstrate that the BWRVIP-108 report results apply to its plant. The five criteria are related to the driving force of the PFM analyses for the recirculation inlet and
- 5 outlet nozzles. The December 19,2007, SE stated that the nozzle material fracture toughness related reference temperature (RT NOT) used in the PFM analyses were based on data from the entire fleet of BWR RPVs. Therefore, the BWRVIP-108 report PFM analyses are bounding with respect to fracture resistance, and only the driving force of the underlying PFM analyses needs to be evaluated. The SE also stated that, except for the RPV heatup/cooldown rate, the plant specific criteria are for the recirculation inlet and outlet nozzles only because the probabilities of failure, P(FIE)s for other nozzles are an order of magnitude lower. The plant-specific heatup/cooldown rate that the NRC staff established in Criterion 1 is regarding the rate under the plant's normal operating condition, which is limiting. This is because after considering the event frequency of 1.0 per reactor year conservatively assumed for the normal operating condition and the event frequency of 1 x1 0-3 per reactor year for a very severe low temperature overpressure (L TOP) transient as determined by the NRC staff in the July 28, 1998, safety evaluation on Topical Report BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," the vessel failure probability for the normal operating condition becomes limiting. The December 19, 2007, safety evaluation concluded that the vessel failure probability of 1.98E-6 for the normal operation is consistent with the NRC's safety goal.
The licensee's submittal conveys NSPM's plant-specific data for the MNGP RPV, and the licensee's evaluation of the five driving force factors, or ratios, against the criteria established in the December 19, 2007, SE. The NRC staff verified the licensee's evaluation, which indicated that all criteria, except for Criterion 4 regarding recirculation outlet nozzles, are satisfied. As a result, the reduced inspection requirements in accordance with ASME Code Case N-702 apply to all proposed MNGP RPV nozzles, which do not include recirculation outlet nozzles; the NRC staff concludes that the licensee's proposed alternative for all MNGP RPV nozzles included in this application (see Section 3.1 of this SE) provides an acceptable level of quality and safety.
It should be noted that the recirculation outlet nozzles are outside the scope of this application because they failed the test of Criterion 4. Further, RPV feedwater nozzles and control rod drive return line nozzles are outside the scope of ASME Code Case N-702 and are, accordingly, outside the scope of this application.
ASME Code Case N-702 permits a VT-1 visual examination of the nozzle inner radius without performing a sensitivity demonstration of detecting a 1-mil width wire or crack. This is not consistent with the NRC position established in Regulatory Guide (RG) 1.147, Revision 15, regarding ASME Code Case N-648-1. However, since the licensee stated in the submittal that "NSPM is not requesting use of the VT-1 examination provisions included in the code case in lieu of performing volumetric examinations," the inconsistency between ASME Code Case N 702 and RG 1.147 regarding VT-1 is not an issue in this application.
4.0 CONCLUSION
The NRC staff has reviewed the licensee submittal regarding the licensee's evaluation of the five plant-specific criteria specified in the December 19, 2007, SE for the BWRVIP-108 report, which provides technical bases for use of ASME Code Case N-702, to examine RPV nozzle-to vessel welds and nozzle inner radii at MNGP. Based on the evaluation in Section 3.2 above, the NRC staff determines that the licensee's proposed alternative provides an acceptable level
-6 of quality and safety and applies to all specified MNGP RPV nozzles. This relief request does not include recirculation outlet nozzles, feedwater nozzles, and control rod drive return nozzles.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR SO.SSa(a)(3)(i), and the licensee is in compliance with the ASME Code's requirements. Therefore, the NRC authorizes the licensee's proposed alternative for inspection of nozzle-to-vessel shell welds and nozzle inner radii sections of RPV nozzles specified in Relief Request RR-002 for MNGP's fifth 10-year lSI interval.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: Simon Sheng, NRR Date: May 22, 2012
- 1. J. O'Connor
- 2 Should you have any questions, please contact Mr. Peter Tam, the MNGP Project Manager, at 301-415-1451.
Sincerely, IRAJ Istvan Frankl, Acting Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosure:
Safety Evaluation cc w/encl: Distribution via ListServ DISTRIBUTION:
PUBLIC RidsOgcRp Resource LPL3-1 Reading RidsAcrsAcnw MailCtr Resource RidsNrrDorlLpl3-1 Resource RidsRgn3MailCenter Resource RidsNrrPMMOIiticelio Resource RidsNrrDeEvib Resource RidsNrrLABTully Resource SSheng. NRR SWiliiams. EDO R-III ADAMS Accession No.: ML12122A980 FICE LPL3-1/PM LPL3-1/LA EVIB/BC*
LPL3-1/BC A ME PTam 05/09/12 BTuII 05/04/12 SRosenber 04/17/12 IFrankl 05/22/12
- Safety evaluation transmitted bye-mail of 4/17/2012 (ADAMS Accession No ML12109A175).
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