ML20275A297
| ML20275A297 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 10/01/2020 |
| From: | Andrew Hon NRC/NRR/DORL/LPL2-2 |
| To: | Sigmon C, Zaremba A Duke Energy Generation Services |
| References | |
| L-2020-LLR-0091 | |
| Download: ML20275A297 (5) | |
Text
From:
Hon, Andrew Sent:
Thursday, October 1, 2020 9:54 AM To:
Zaremba, Arthur H.; Sigmon, Chet Austin
Subject:
Request for additional information - Brunswick Request for Alternate Examination of Reactor Vessel Nozzles (EPID: L-2020-LLR-0091)
By letter dated June 23, 2020 (Agencywide Documents and Access Management System (ADAMS)
Accession No. ML20181A004. Reference 1), as supplemented by letter dated July 30, 2020 (ADAMS Accession No. ML20212L731), Duke Energy (the licensee) submitted relief request (RR) RA-19-0447 in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(1)(ii) to the requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) at Brunswick Steam Electric Plant, Unit 1. Specifically, you submitted for the U. S. Nuclear Regulatory Commission (NRC) review a request to approve examination relief on the basis that the proposed alternative provides an acceptable level of quality and safety, in accordance with 10 CFR 50.55a(z)(1). The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The proposed questions were discussed by telephone with your team on September 30, 2020.
Your team confirmed that the request for additional information (RAI) was understood, it does not contain the proprietary information, and agreed to provide a response by October 30, 2020.
REGULATORY BASIS The regulations in 10 CFR 50.55a(g) require that the ISI of ASME Code Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Code and applicable addenda. The ASME Code,Section XI, requires that all reactor vessel nozzles to be inspected during each 10-year ISI interval. The volumes in each nozzle required to be inspected are 100 percent of the nozzle-to-vessel shell weld volume and 100 percent of the nozzle inner radius section volume, as shown in the applicable figure in Figures IWB-2500-7(a) through (d) "Nozzle in Shell or Head," of the ASME Code,Section XI.
10 CFR 50 Appendix H Paragraph III.A, requires that no material surveillance program is required for reactor vessels for which it can be conservatively demonstrated by analytical methods applied to experimental data and tests performed on comparable vessels, making appropriate allowances for all uncertainties in the measurements, that the peak neutron fluence at the end of the design life of the vessel will not exceed 1017 n/cm2 (E 1 MeV. NRC Regulatory Guide (RG) 1.190 (Reference 2) describes the NRC approved methodology to calculate the neutron fluence.
REQUESTS FOR ADDITIONAL INFORMATION
RAI 1
Issue:
The licensee states in Section 4 of Enclosure 1 to the supplement dated July 30, 2020 that 2 million realizations were performed as part of the probabilistic fracture mechanics (PFM) analysis (see fifth paragraph of Section 4). However, the licensee states in Section 5 of Enclosure 1 to the Supplement dated July 30, 2020 that no failures occurred for any path in 1 million simulations A similar statement is found in Enclosure 4 to letter dated June 23, 2020. The staff takes the term simulation to
mean realization, in this context. There is an apparent discrepancy in the submitted documents on the number of realizations performed as part of the PFM analysis.
Request:
Clarify how many realizations were performed as part of the PFM analysis.
RAI 2
Issue:
The licensee states in Section 4 of Enclosure 1 to the supplement dated July 30, 2020 that the probability of failure is estimated as 1 failure / 1 million realizations / 60 years = 1.67 x 10-8 per year. This calculation implies that a converged solution was reached in the PFM. The uncertainty in the mean failure probability is not addressed in the licensees PFM analysis. This uncertainty may be important when comparing the mean failure probability to the chosen acceptance criterion of 5x10-6 per year, if the acceptance criterion is within two standard deviations of the mean failure probability.
Request:
Provide (1) a discussion of the uncertainty on the mean failure probability in relation to the acceptance criterion and (2) a discussion of solution convergence.
RAI 3
Issue:
On page 6 of Enclosure 1 to the letter dated June 23, 2020, the licensee states that they will perform either a volumetric exam or VT-1 and that the VT-1 examination is outlined in Code Case N-648-2, Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles. The staff notes that the NRCs condition on Code Case N-702 in Regulatory Guide 1.147, Revision 19 requires the use of Code Case N-648-2 if VT-1 is used in place of the volumetric exam.
Request:
Confirm that Code Case N-648-2 will be used for VT-1 exams, in accordance with NRCs condition on Code Case N-702.
RAI 4
Issue:
The staff noted that the probability of failure at the blend radius of the Brunswick recirculation outlet nozzle due to low temperature overpressure (LTOP), as shown in Table 12 of Enclosure 1 of the July 30, 2020 supplement, is much lower compared to the probability of failure at the blend radius in one of the nozzles (Columbia) analyzed in BWRVIP-241, BWR Vessel and Internals Project: Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii. The staff noted that the transient cycles for the Brunswick recirculation outlet nozzle (Table 5 of to the supplement) are similar to the transient cycles for the Columbia recirculation outlet nozzle (Table 5-5 of BWRVIP-241). The staff also compared stresses and noted that the stresses at the Brunswick recirculation outlet nozzle blend radius are higher (Figures 10 and 11 of Enclosure 1 to the supplement) than those at the Columbia recirculation outlet nozzle blend radius (Figures 4-44 through 4-47 of BWRVIP-241). Given that the random parameters are the same for both cases (Table 11 of to the supplement for Brunswick and Table 5-1 of BWRVIP-241, Case 5, for Columbia), the
staff expected that the probability of failure for Brunswick would be slightly higher than the probability of failure for Columbia. However, the LTOP probability of failure at the blend radius for Brunswick is five orders of magnitude lower than for the corresponding case for Columbia (Table 12 of Enclosure 1 to the supplement for Brunswick compared to Table 5-9 of BWRVIP-241 for Columbia).
Request:
Explain why the probability of failure at the nozzle blend radius of the Brunswick recirculation outlet nozzle due to LTOP is so much lower than the corresponding probability of failure for the Columbia recirculation outlet nozzle.
RAI 5
Issue:
The licensee stated that the neutron fluence projections or evaluation for the BSEP reactor pressure vessel nozzle-to-vessel welds and inner radii over the period of extended operation (54 effective full power years) can be found from RR, WCAP-17660 (Reference 3) and Pressure-Temperature Limits report elated documents. The staff uses the neutron fluence values as reported in WCAP-17660 to infer the neutron fluence at the recirculation outlet nozzles below:
Neutron Fluence Projections at 54 EFPY (WCAP-17660)
Unit 1 Unit 2 Girth Weld FG (254 AVO*)
1.0x1018 ~ 2.8x1018 n/cm2 (Table 2.2-17) 0.9x1018 ~ 2.9x1018 n/cm2 (Table 2.2-42)
H6A (183 AVO) 4.7x1017 ~ 1.9x1018 n/cm2 (Table 2.2-6) 4.7x1017 ~ 2.0x1018 n/cm2 (Table 2.2-31)
H6B (179 AVO) 1.8x1017 ~ 7.2x1017 n/cm2 (Table 2.2-7) 1.8x1017 ~ 7.5x1017 n/cm2 (Table 2.2-32)
Girth Weld GH (110 AVO) 0.8x1012 ~ 2.3x1012 n/cm2 (Table 2.2-16) 0.8x1012 ~ 2.3x1012 n/cm2 (Table 2.2-41)
- AVO = Above Vessel Zero Based on Figure 2.1-4 of WCAP-17660, the recirculation outlet nozzles are located between the elevations of Girth Weld GH and Girth Weld FG. The actual elevation for recirculation outlet nozzle center is 161 AVO (Reference 4). By adding up the outer radius of the nozzle, 24, to the center elevation, it appears that the upper 25% of the nozzle would be exposed to a neutron fluence > 1.0x1017 n/cm2 at 54 EFPY.
Request:
Provide additional information or justification for the conclusion that the neutron fluence for the welds between the BSEP reactor pressure vessel and nozzles is less than 1.0x1017 n/cm2 at 54 EFPY.
References
- 1. Duke Energy to NRC, Proposed Alternative for RPV Nozzle-to-Vessel Weld and Inner Radii Examination Requirements in Accordance with 10 CFR 50.55a(z)(1), June 23, 2020. (ADAMS Accession No. ML20181A004).
- 2. Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.
- 3. WCAP-17660-NP, Revision 0, Neutron Exposure Evaluations for Core Shroud and Pressure Vessel Brunswick Units 1 and 2, November 2012.
- 4. Vendor Drawing VPF-2478-22-9, U1 & 2 Reactor Nozzle Location The NRC staff considers that timely responses to RAIs help ensure sufficient time is available for staff review and contribute toward the NRCs goal of efficient and effective use of staff resources. Please note that if you do not respond to this request by the agreed-upon date or provide an acceptable alternate date, we may deny your application for amendment under the provisions of Title 10 of the Code of Federal Regulations, Section 2.108. If circumstances result in the need to revise the agreed upon response date, please contact me.
Andy Hon, PE Project Manager (Brunswick Nuclear Plant 1 & 2, Duke Energy Fleet)
Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 301-415-8480 OWFN O8E06 Mail Stop O8B1A andrew.hon@nrc.gov
Hearing Identifier:
NRR_DRMA Email Number:
815 Mail Envelope Properties (SA0PR09MB677949938280F142689E384999300)
Subject:
Request for additional information - Brunswick Request for Alternate Examination of Reactor Vessel Nozzles (EPID: L-2020-LLR-0091)
Sent Date:
10/1/2020 9:54:08 AM Received Date:
10/1/2020 9:54:08 AM From:
Hon, Andrew Created By:
Andrew.Hon@nrc.gov Recipients:
"Zaremba, Arthur H." <Arthur.Zaremba@duke-energy.com>
Tracking Status: None "Sigmon, Chet Austin" <Chet.Sigmon@duke-energy.com>
Tracking Status: None Post Office:
SA0PR09MB6779.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 9621 10/1/2020 9:54:08 AM Options Priority:
Normal Return Notification:
No Reply Requested:
No Sensitivity:
Normal Expiration Date: