ML24054A158
| ML24054A158 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 02/27/2024 |
| From: | Mary Johnson NRC/NRR/DNRL/NLRP |
| To: | Hafen S Northern States Power Company, Minnesota |
| References | |
| EPID L-2022-SLR-0000 | |
| Download: ML24054A158 (11) | |
Text
February 27, 2024 Mr. Shawn Hafen Site Vice President Northern States Power Company -
Minnesota Monticello Nuclear Generating Plant 2807 West County Road 75 Monticello, MN 55362
SUBJECT:
MONTICELLO NUCLEAR GENERATING PLANT - LIMITED AGING MANAGEMENT AUDIT REPORT REGARDING THE SUBSEQUENT LICENSE RENEWAL APPLICATION REVIEW (EPID NO. L-2022-SLR-0000)
Dear Mr. Hafen:
By letter dated January 9, 2023 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML23009A352), as supplemented by letters dated April 3, 2023 (ML23094A136), June 26, 2023 (ML23177A218), July 11, 2023 (ML23193B026), July 18, 2023 (ML23199A154), August 15, 2023 (ML23227A175), August 28, 2023 (ML23240A695), September 5, 2023 (ML23248A474), September 22, 2023 (ML23265A158), October 3, 2023 (ML23276B433), and November 9, 2023 (ML23313A158), Northern States Power Company, Minnesota submitted to the U.S.
Nuclear Regulatory Commission (NRC or staff) an application to renew the Renewed Facility Operating License No. DPR-22 for Monticello Nuclear Generating Plant (MNGP).
S. Hafen 2
The NRC conducted a limited aging management audit from October 19, 2023 - January 4, 2024, in accordance with the audit plan (Package Accession No. ML23289A144). The audit was extended beyond the originally planned date to allow for additional interactions between NRC and MNGP staff members. The audit report is enclosed. If you have any questions on this matter, please contact Marieliz Johnson, Project Manager, for the safety review of the MNGP Subsequent License Renewal Application at Marieliz.Johnson@nrc.gov, Sincerely,
/RA/
Marieliz Johnson, Project Manager License Renewal Projects Branch Division of New and Renewed Licenses Office of Nuclear Reactor Regulation Docket No. 50-263
Enclosure:
Audit Report cc: w/encl.: ListServ
- via email NRR-106 OFFICE NRR/DNRL: PM NRR/DNRL:LA NRR/DEX: BC NAME MJohnson KBratcher ITseng DATE 02/23/2024 02/26/2024 02/13/2024 OFFICE NRR/DNRL: BC NRR/DNRL: BC NAME ABuford LGibson DATE 02/13/2024 02/27/2024
AUDIT REPORT Aging Management Audit Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application October 19, 2023 - January 4, 2024 Division of New and Renewed Licenses Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION DIVISION OF NEW AND RENEWED LICENSES Docket No:
50-263 License No:
DPR-22 Licensee:
Northern States Power Company, a Minnesota corporation (NSPM)
Facility:
Monticello Nuclear Generating Plant (MNGP)
Location:
Rockville, Maryland Dates:
October 19, 2023 - January 4, 2024 Approved By:
Ian Tseng, Chief Structural, Civil, Geotech Engineering Branch Division Of Engineering and External Hazards Angie Buford Vessels and Internals Branch Division of New and Renewed Licenses Lauren K. Gibson, Chief License Renewal Projects Branch Division of New and Renewed Licenses Reviewers:
Name Organization George Thomas Andrew Prinaris David Dijamco Structural, Civil, Geotech Engineering Branch (ESEB)/ Division of Engineering and External Hazards (DEX)
ESEB/DEX Vessels and Internals Branch (NVIB)/ New and Renewed Licenses (DNRL)
Marieliz Johnson License Renewal Projects Branch (NLRP)/DNRL
TABLE OF CONTENTS AUDIT REPORT......................................................................................................................
TABLE OF CONTENTS...........................................................................................................
- 1. Background......................................................................................................................
- 2. Audit Activities..................................................................................................................
- 3. Documents Reviewed.......................................................................................................
- 4. Audit Observations:..........................................................................................................
- 5. Supplements to the SLRA................................................................................................
- 6. List of Participants............................................................................................................
- 7. Exit Meeting......................................................................................................................
Report for the Aging Management Audit Monticello Nuclear Generating Plant, Unit 1 Subsequent License Renewal Application Aging Management Regulatory Audit
- 1.
Background
By letter dated January 9, 2023 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML23009A352), as supplemented by letters dated April 3, 2023 (ML23094A136), June 26, 2023 (ML23177A218), July 11, 2023 (ML23193B026), July 18, 2023 (ML23199A154), August 15, 2023 (ML23227A175), August 28, 2023 (ML23240A695), September 5, 2023 (ML23248A474), September 22, 2023 (ML23265A158), October 3, 2023 (ML23276B433), November 9, 2023 (ML23313A158),
and November 30, 2023 (ML23334A147), Northern States Power Company, Minnesota (NSPM or the applicant) submitted to the U.S. Nuclear Regulatory Commission (NRC or staff) an application to renew the Renewed Facility Operating License No. DPR-22 for Monticello Nuclear Generating Plant (MNGP). The staff is reviewing this application in accordance with the guidance in NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, (SRP-SLR).
Following review of Xcel Energy Subsequent License Renewal Application (SLRA) Supplement 5 dated August 28, 2023 (ML23240A695), as related SLRA Section 3.5.2.2.2.6, Reduction of Strength and Mechanical Properties of Concrete Due to Irradiation, the staff will conduct a limited audit to review the applicants evaluations and supporting bases for its conclusions in the SLRA supplement related to the following topics: (1) Evaluation of gamma heating and gamma dose effects on biological shield concrete, (2) proposed AMR items for managing loss of fracture toughness of Reactor Vessel (RV) steel support and biological shield structural steel, and (3) Evaluation of loss of fracture toughness of biological shield structural steel.
- 2.
Audit Activities The NRC staff performed a limited audit specific to SLRA Further Evaluation Section 3.5.2.2.2.6 supplements. During this limited audit, the staff reviewed SLRA Section 3.5.2.2.2.6, as modified by the following Enclosures of SLRA Supplement 5 dated August 28, 2023 (ML23240A695):
Enclosures 01, Irradiation Effects on Biological Shield Structural Steel Components Enclosures 02, Loss of Fracture Toughness of RV Supports
- 3a, Reduction of Strength and Mechanical Properties of Structural Steel and Concrete Due to Irradiation
- 3c, Gamma Dose Biological Shield Irradiation Evaluation Clarification
- 3d, Reactor Vessel Support Steel Irradiation Evaluation Clarifications
- 3e, Biological Shield Structural Steel Evaluation Clarifications
- 4, Concrete Aging Management Review - Containment Temperature Control Clarification.
The staff also reviewed the related AMR items in Enclosure 3 of Supplement 7 dated November 30, 2023 (ML23334A147).
Specifically, the staff reviewed the applicants evaluations and supporting bases for its conclusions in the SLRA Supplement 5 related to the following topics: (1) Evaluation of gamma heating and gamma dose effects on biological shield concrete, (2) proposed AMR items for managing loss of fracture toughness of RV steel support and biological shield structural steel, and (3) Evaluation of loss of fracture toughness of biological shield structural steel.
During its audit, the NRC staff discussed with the applicant and reviewed documentation contained in the SLRA Supplement 5 Enclosures 01, 02, 03a, 03c, 03d, 03e and 04, as well as Supplement 7 Enclosure 3, and those in applicants ePortal. The table below lists documents that were reviewed by the NRC staff and were found relevant to the SLRA Section 3.5.2.2.2.6, as modified by SLRA Supplements 5 and Supplement 7. The staff will document its review of this information in the SER.
- 3.
Documents Reviewed Document Title Revision / Date Biological Shield Wall (BSW) Concrete -
Reduction of Strength:
NSPM Calculation 23-016 (Structural Integrity Associates (SIA)
Calculation 2200285.401)
Heat Transfer Analysis of MNGP Bioshield Revision 0 (08/24/2023)
Calculation MN09-995-091-100 Determination of Concrete Strength Using Test Data, Attachment A
Revision 0 N/A Concrete Placement Checklist - Reactor Building, Pour No. 381, EL 947.2 to 963, Concrete Mix C2-P, 02/03/1969 126413 Concrete Reactor Superstructure - Concrete Placement and Test Report, Monticello NPP, Bechtel Corporation N/A N/A Weekly Schedule Meeting (includes Pour Schedule for Sacrificial Shield) 04/02/1969 CA-97-007 Calculation of Local Temperature in Biological Shield Wall Revision 0 901/02/1997)
SIA Calculation Package 2200285.301 Monticello SLR Concrete Embrittlement Assessment of Biological Shield Wall, Evaluation of Concrete Degradation of MNGP Bioshield, Structural Integrity Associates, Inc (SIA).
Revision 0 (10/05/2022)
RIL 2021-07 (ANL/EVS-20/18)
Radiation Effects on Concrete - An Approach for Modeling Degradation of Concrete Properties August 2021 ORNL/TM-2018/769 Expected Condition of Concrete Exposed to Radiation at Age 80 Years of Reactor Operation, Light Water Reactor Sustainability Program, US Department of Energy Revision 0 (January 2018)
N/A Bruck et al., Structural assessment of radiation damage in light water power reactor concrete biological shield walls, Nuclear Engineering and Design, Volume 350, pages 9-20 08/15/2019 N/A MNGP Presentation-Bioshield Analysis Conservatism N/A N/A Hilsdorf, H.K., Kropp, J., and Koch, H.J, The Effects of Nuclear Radiation on the Mechanical Properties of Concrete, American Concrete Institute SP 55-10, p223-234 1978 BSW Steel Embrittlement NSPM Calculation 23-017 (SIA Calculation 2200285.402)
MNGP Bioshield Transition Temperature Approach and Fracture Mechanics per NUREG-1509 Guidance Revision 0 (08/24/2023)
N/A Maxson Corporation Certified Mill Test Reports - Monticello NPP for A36 structural steel 01/17/1969 5828-C-37 Specification for Detailing, Fabrication, Delivery and Erection of Structural Steel for the Reactor Building - Drywell Interior of MNGP Unit 1 08/25/1967 NF-36351 Drawing - Reactor Building Drywell and Interior Structural Steel &
Liner Plates for Biological Shield Revision 76 (08/17/2009)
S8 Encl 03 - TRP 76-S14 Item 1.pdf Responses to select audit questions regarding fracture mechanics evaluation of bioshield structural steel N/A Bioshield conservatism with notes.pdf Slide presentation of conservatisms used in the bioshield evaluation N/A
- 4.
Audit Observations:
Bioshield Wall Concrete Embrittlement:
The generic curve from literature for the variation of gamma flux/dose normalized to flux at core-midplane along the height of the BSW concrete used in the SLRA calculation SIA 2200285.301 for attenuation of gamma dose is from Figure 2-7 of EPRI Report 3002011710, Irradiation Damage of the Concrete Biological Shield -
Basis for Evaluation of Concrete Biological Wall for Aging Management, May 2018.
The staff noted that the same curve appears as Figure 3-7 Axial Gamma Flux Variation Relative to Core Fuel Mid-Height of open-source Report ORNL/TM-2018/769, Revision 0, Expected Condition of Concrete Exposed to Radiation at Age 80 Years of Reactor Operation, US Department of Energy, January 2018. The applicant may need to reflect this information in the SLRA. The staff also noted from Figure 4 of the audited calculation 2200285.301 that using the above referenced curve, the normalized gamma flux was extrapolated to a value of zero at a distance approximately 100 inches (8 ft 4 in) from the fuel core mid-height.
In calculation SIA 2200285.301 the applicant determined that the reduction in specified concrete compressive strength (fc = 4,000 psi) due to gamma radiation was based on data from the paper by Hilsdorf, H.K., Kropp, J., and Koch, H.J, The Effects of Nuclear Radiation on the Mechanical Properties of Concrete, American Concrete Institute SP 55-10, p223-234, 1978. The staff noted that a value of reduced concrete compressive strength of approximately 0.9 x fc = 3,600 psi, further rounded down to 3,500 psi was used for re-evaluation of the BSW structural concrete portion.
The staff also noted that bounding loads and acceptance criteria used in the re-evaluation of the BSW were based on original design calculations. The applicant may need to state this information in the SLRA.
Audited plant-specific concrete test data of concrete pours from original construction representative of the BSW wall concrete indicate that the measured (tested) compressive strengths at 28 days and 90 days generally exceeded 4,600 psi (i.e.,
1.15 times the specified compressive strength, fc). The applicant may need to state this information in the SLRA.
NSPM Calculation 23-016 shows that from a heat transfer analysis through the reactor vessel air cavity gap, the upper bound temperature value of the BSW steel liner and concrete inner surfaces was estimated to be 140.69oF.
NSPM Calculation 23-016 did not include a plant-specific analysis of temperature increase into the concrete due to thermal conductivity and gamma heating. Instead, the staff noted that the applicant did an extrapolation of calculated results shown in case study Figure 5-6 of a meso-scale modeling approach described in NRC Research Information Letter (RIL) 2021-07 to estimate the maximum temperature increase due to gamma heating only, as 1.12oF. The staff noted that applicability of the assumed (hypothetical) input properties and parameters used in the RIL case study meso-scale model (listed in RIL Tables 2-3, 4-1 and 5-1) were not plant-specific to MNGP concrete at the mesoscale. The applicant may need to further justify the applicability of the RIL used parameters and also address increase in temperature inside the concrete due to thermal conductivity in the SLRA.
BSW Steel Embrittlement:
Appendix A of NSPM Calculation 23-016 (SIA 2200285.401) included a detailed heat transfer and BSW concrete expansion evaluation that included calculations of the stress acting on the inner and outer BSW steel liners due the concrete-steel differential thermal expansion. The resulting hoop stress in the inner steel liner due to concrete-steel differential thermal expansion is a compressive stress of 0.936 ksi.
The resulting hoop stress in the outer steel liner due to concrete-steel differential thermal expansion is a tensile stress of 0.687 ksi.
NSPM Calculation 23-017 (SIA 2200285.402) included details of the fracture mechanics evaluation of the BSW structural steel.
File S8 Encl 03 - TRP 76-S14 Item 1.pdf included clarifications on the fracture mechanics evaluation of the BSW structural steel.
NSPM Calculation 23-017 (SIA 2200285.402) states that the bounding lower temperature for operating temperature is 111.91°F, which is a plant-specific value and is at the outermost distance from the bioshield steel liner.
Figure 6 of NSPM Calculation 23-016 (SIA 2200285.401) shows that the temperature at the radial location of inner steel liner is approximately 120°F at the mid-core level.
Figure 3 of NSPM Calculation 23-017 (SIA 2200285.402) shows that the maximum stress location is at the inner steel liner.
The Certified Mill Test Reports for the Monticello BSW structural steel did not include initial nil ductility temperatures.
The staff will consider issuing Requests for Additional Information (RAIs) to obtain the necessary information and/or use a voluntary SLRA supplement offered by the applicant to address issues identified and communicated during the audit. The staff will document its review of this information and resolution of the issues in the safety evaluation.
- 5.
Supplements to the SLRA By Subsequent License Renewal Application Supplement 8|letter dated January 11, 2024]] (ML24012A051), NSPM voluntarily submitted Supplement 8 to the SLRA resulting from discussions held during the audit.
- 6.
List of Participants Xcel Energy Paul Young, Manager of Projects Max Smith, Project Manager Steve Sollom, Senior Engineer Matt Sears, Senior Engineer Toutseng Hawj, Senior Engineer Jason Tribe, Senior Engineer Enercon Jeff Gromatzky, License Renewal Supervisor Reene Gambrell, License Renewal Lead Ian Miller, License Renewal Supervisor Mitch McFarland License Renewal Senior Engineer Structural Integrity Associates (SIA)
Dan Denis, Vessel & Internals Manager
- 7.
Exit Meeting An exit meeting was held with the applicant on January 4, 2024, to discuss the results of the regulatory audit.