ML24113A127
| ML24113A127 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/22/2024 |
| From: | Joel Wiebe NRC/NRR/DORL/LPL3 |
| To: | Rhoades D Constellation Energy Generation |
| Shared Package | |
| ML24113A124 | List: |
| References | |
| EPID L-2023-LLA-0136 | |
| Download: ML24113A127 (12) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION April 22, 2024 David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNIT NOS. 1 AND 2 - AUDIT PLAN IN SUPPORT OF REVIEW OF LICENSE AMENDMENT REQUEST REGARDING REVISION OF TECHNICAL SPECIFICATIONS 3.7.15, "SPENT FUEL POOL BORON CONCENTRATION," 3.7.16, "SPENT FUEL ASSEMBLY STORAGE, AND 4.3.1 "FUEL STORAGE, CRITICALITY" (EPID L-2023-LLA-0136)
Dear David P. Rhoades:
By letter dated September 29, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23272A201), Constellation Energy Generation, LLC (Constellation, the licensee), submitted a license amendment request (LAR) for Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2, to the U.S. Nuclear Regulatory Commission (NRC).
Specifically, the proposed LAR would modify several technical specifications related to spent fuel pool boron concentration, criticality, and spent fuel assembly storage.
During the review of the LAR, the NRC staff identified items that require further clarification and detailed explanations. The NRC staff will conduct a regulatory audit to support its review of the LAR in accordance with the enclosed audit plan. A regulatory audit is a planned activity that includes the examination and evaluation of primarily non-docketed information. The audit will be conducted to increase the NRC staffs understanding of the LAR and may identify information that will require docketing to support the NRC staffs regulatory findings.
The audit will be conducted using an online portal and teleconferences. The audit plan and supporting materials are enclosed.
to this letter contains proprietary information. When separated from, this document is DECONTROLLED.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION If you have any questions, please contact me by telephone at 301-415-6606 or via email at Joel.Wiebe@nrc.gov.
Sincerely,
/RA/
Joel S. Wiebe, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-454, STN 50-455, STN 50-456, and STN 50-457
Enclosures:
- 1. Proprietary Audit Plan
- 2. Nonproprietary Audit Plan cc: Listserv
ENCLOSURE 2 REGULATORY AUDIT PLAN REGARDING LICENSE AMENDMENT REQUEST TO MODIFY TECHNICAL SPECIFICATIONS 3.7.15, "SPENT FUEL POOL BORON CONCENTRATION," 3.7.16, "SPENT FUEL ASSEMBLY STORAGE, AND 4.3.1 "FUEL STORAGE, CRITICALITY" BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNIT NOS. 1 AND 2 CONSTELLATION ENERGY GENERATION, LLC.
DOCKET NOS. STN 50-454, STN 50-455, STN 50-456, AND STN 50-457 (NON-PROPRIETARY)
Proprietary information pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations has been redacted from this document.
Redacted Proprietary information is identified by empty double brackets.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION REGULATORY AUDIT PLAN REGARDING LICENSE AMENDMENT REQUEST TO MODIFY TECHNICAL SPECIFICATIONS 3.7.15, "SPENT FUEL POOL BORON CONCENTRATION," 3.7.16, "SPENT FUEL ASSEMBLY STORAGE, AND 4.3.1 "FUEL STORAGE, CRITICALITY" BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNIT NOS. 1 AND 2 CONSTELLATION ENERGY GENERATION, LLC.
DOCKET NOS. STN 50-454, STN 50-455, STN 50-456, AND STN 50-457
1.0 BACKGROUND
By letter dated September 29, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23272A201), Constellation Energy Generation, LLC (Constellation, the licensee), submitted a license amendment request (LAR) for Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2, to the U.S. Nuclear Regulatory Commission (NRC).
The proposed amendment revises Technical Specifications 3.7.15, "Spent Fuel Pool Boron Concentration," 3.7.16, "Spent Fuel Assembly Storage, and 4.3.1 "Fuel Storage, Criticality."
2.0 REGULATORY AUDIT BASES A regulatory audit is a planned license or regulation-related activity that includes the examination and evaluation of primarily non-docketed information. The audit is conducted with the intent to gain understanding, to verify information, and to identify information that will require docketing to support the basis of a licensing or regulatory decision. Performing a regulatory audit is expected to assist the NRC staff in efficiently conducting its review and gaining insights to the licensees processes and procedures. Information that the NRC staff relies upon to make the safety determination must be submitted on the docket.
This regulatory audit is based on the following regulatory requirements and guidance:
Title 10 of the Code of Federal Regulations (10 CFR), part 50, appendix A, General Design Criterion (GDC) 44, Cooling Water, GDC 61, Fuel Storage and Handling and Radioactivity Control, GDC 61, Prevention of Criticality in Fuel Storage and Handling, 10 CFR 50.68, Criticality Accident Requirements,10 CFR 50.36, Technical Specifications,
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Regulatory Guide (RG) 1.240, Revision 0, Fresh and Spent Fuel Pool Criticality Analysis, RG 1.13, Revision 2, Spent Fuel Pool Storage Facility Design Basis, and NUREG 0800, Standard Review Plan (SRP) 9.1.3, Spent Fuel Pool Cooling and Cleanup System.
NUREG 0800, SRP 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling.
NUREG 0800, SRP 9.1.2, New and Spent Fuel Storage.
3.0 SCOPE The audit team will view the documentation and calculations that provide the technical support for the LAR. The scope of the NRC staffs audit will focus on the following subjects:
Calculations and analyses supporting the spent fuel pool heat up, Calculations and analyses supporting fuel decay heat assumptions, Significance of different core offload cases used, Calculations and analyses supporting the spent fuel pool subcriticality, and Calculations and analyses supporting structural integrity of the spent fuel pool.
In addition, the audit team will request to discuss these topics with Constellations subject matter experts. The NRC staff will conduct this audit under the guidance provided in NRR Office Instruction LIC-111, Regulatory Audits, Revision 1 (ML19226A274).
4.0 INFORMATION AND OTHER MATERIAL NECESSARY FOR THE REGULATORY AUDIT The NRC staff requests that the documents, data, and calculations regarding the following topics be made available to the staff in Constellations electronic reading room, specifically, information related to the following and any supporting documents:
No.
LAR Ref. No.
Document Title 1
6.1.9 HI-982094, Revision 5, "Criticality Evaluation for the Byron/
Braidwood Rack Installation Project," dated December 2013 2
6.2.3 BRW-22-0026-N Revision 0 and BYR22-018 Revision 0, "Criticality Calculation Benchmarking with MCNP V6.2,"
dated August 2023 3
6.2.4 BRW-22-0027-N Revision 0 and BYR22-019 Revision 0, "Braidwood-Byron Criticality Analysis for New Fuel Vault,"
dated August 2023.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION No.
LAR Ref. No.
Document Title 4
6.2.5 BRW-22-0028-N Revision 0 and BYR22-020 Revision 0, "Braidwood-Byron Criticality Analysis for Spent Fuel Pool,"
dated September 2023 5
Braidwood TS 3.7.16 LCO a and b both say of Holtec spent fuel pool storage racks. Include information related to the specificity of including the label Holtec SFP racks in the Braidwood TS 3.7.16 LCO as well as the Braidwood TS 4.3.1.c and d?
6 Braidwood TS 3.7.16 SR 3.7.16.2 includes decay time as a factor in Figure 3.7.16-1. The proposed replacement Figure 3.7.16-1 does not include any basis on decay time. Include information related to why isnt decay time being removed from Braidwood TS 3.7.16 SR 3.7.16.2?
7 Fuel enrichments above 5.0 wt/% U-235 are outside the typical geometries and compositions associated with spent fuel pools and bundle designs that were in widespread use in the United States when the guidance in NEI 12-16 and Regulatory Guide 1.240 was developed. Per Regulatory Guide 1.240, paragraph C.1.o analysis of such fuel using the guidance may require additional justification to use the guidance. Include information related to justification for applying the guidance to the analysis of fuel greater than 5.0 wt/% U-235.
8 Per the NRCs approval of CASMO5 in Studsvik topical report SSP-14-P01/028-TR-P-A, Generic Application of the Studsvik Scandpower Core Management System to Pressurized Water Reactors, Revision 0 (ADAMS ML17279A985) CASMO5 is limited to U-235 enrichments of 5.0 wt/% U-235 and below. Include information related to the licensees use of CASMO5 to include:
Its use outside the scope of its approval and the overall effect on the LAR.
Confirm that CASMO5 was used in accordance with topical report SSP-14-P01/028-TR-P-A.
Identify and remove any impact on the LAR by analysis of fuel with U-235 enrichments above 5.0 wt/% by CASMO5, i.e. the burnup/enrichment loading curve. The third order polynomial in Table 8 includes data from above 5.0 wt/%
Identify the code to be used to determine a fuel assemblys burnup and whether there is any bias and/or uncertainty between the burnup calculated by that code and CASMO5, and if there is that bias and/or uncertainty must be incorporated into the analysis.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION No.
LAR Ref. No.
Document Title 9
The Byron and Braidwood SFP capacity. TS capacity for each is 2984, yet actual capacity as stated in the LAR is 2588/2568 cells at Byron and Braidwood respectively.
The difference is large. Include information related to the delta between the actual and licensed capacity and the potential impact on the SFP NCS analysis and subsequent requirements.
10 Include information related to the licensees as built documentation for its SFPs and the SFP models used in the analysis.
11 The licensees analysis treats ((
)). Include information related to the licensees justification for this deviation from the Guidance in NEI 12-16 R4/Regulatory Guide 1.240 R0.
12 In the LAR the licensee stated, The criticality safety analysis of record (Reference 6.1.9) [HI-982094, Revision 5, "Criticality Evaluation for the Byron/Braidwood Rack Installation Project", dated December 2013.] for the SFP will still cover Westinghouse supplied 17x17 fuel assemblies (OFA (Optimized Fuel Assembly), VANTAGE 5, VANTAGE+) and previously irradiated Westinghouse and Framatome Lead Use Assemblies (LUAs). The enrichment limit for Westinghouse supplied 17x17 fuel assemblies (OFA, VANTAGE 5, VANTAGE+) and Framatome LUAs is 5.0 weight percent (wt%) Uranium 235 (U-235). The Byron and Braidwood TS 4.3.1.b indicate Holtec International Report HI-982094, "Criticality Evaluation for the Byron/Braidwood Rack Installation Project", Project 80944 1998 as the analysis of record. Include information related to why there is a difference between the analysis of record and the TS?
13 In the LAR the licensee stated, The criticality safety analysis of record (Reference 6.1.9) [HI-982094, Revision 5, "Criticality Evaluation for the Byron/Braidwood Rack Installation Project", dated December 2013.] for the SFP will still cover Westinghouse supplied 17x17 fuel assemblies (OFA (Optimized Fuel Assembly), VANTAGE 5, VANTAGE+) and previously irradiated Westinghouse and Framatome Lead Use Assemblies (LUAs). The enrichment limit for Westinghouse supplied 17x17 fuel assemblies (OFA, VANTAGE 5, VANTAGE+) and Framatome LUAs is 5.0 weight percent (wt%) Uranium 235 (U-235). What TS or license controls prevents storage of Westinghouse Vantage 5, Vantage 5+, and OFA above 5.0 wt%?
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION No.
LAR Ref. No.
Document Title 14 The licensee included a fuel depletion related geometry change bias. Include information related to the details about how the bias was determined and the extent to which it is limited by fuel type and material as well as the justification for using the bias.
15 The licensees treatment of fuel assembly manufacturing tolerance and uncertainties are generally consistent with the Guidance in NEI 12-16 R4/Regulatory Guide 1.240 R0, except the licensee did not include an ((
)). Include information related to ((
)). Include information related to ((
)) was modeled in the SFP accident analysis.
16 Include information related to the list of isotopes that was modeled in SFP criticality analysis.
17 In the LAR the licensee states, Depletion calculations were performed with conservative operating conditions: highest fuel temperature, moderator temperature and soluble boron concentrations during in-core operation. Include information related to the details about what exactly was used, how it compares to past operation and how it bounds future operation with the fuel changes and transition to 24-month cycles.
18 With respect to integral and fixed burnable absorbers, in the LAR the licensee states, Gadolinia burnable absorbers were also conservatively neglected. Include information related to the justification for why this was conservative for the plants and planned operations.
19 Include information related to control element assembly usage during operation or justification for not including in the analysis.
20 The licensee included a fission gas release bias. However, the licensee did not provide any details about how the bias was determined or the extent to which it is limited by fuel type and material. Include information related to the justification for the fission gas release bias.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION No.
LAR Ref. No.
Document Title 21 In the LAR the licensee states, In the case that any fuel assembly needs fuel reconstitution, the activity will be performed with the assembly isolated from any other fuel assembly. Further evaluation will be performed for the reconstituted fuel assemblies separately. This is directed by Constellation Procedure NF-AP-309, "PWR Special Nuclear Material and Core Component Move Sheet Development.
However, the licensee did not specify what constitutes
...isolated from any other fuel assembly. Nor did the licensee describe the methodology that would be used to evaluate reconstituted/non-standard fuel assemblies.
Constellation procedure NF-AP-309 was not submitted as part of the LAR and hence is unable to be reviewed by the NRC staff. Include information related to the justification for the treatment of reconstituted/non-standard fuel assemblies.
22 In the LAR the licensee provides information on several non-conforming fuel assemblies and their disposition. However, the licensee did not specify how the disposition would be maintained in perpetuity. Include information related to the justification for the treatment of these non-conforming fuel assemblies.
23 In the LAR the licensee credits administrative control and process check to preclude analyzing misloading multiple fresh assemblies into Region 2. However, the LAR does not state what administrative control and process check are being relied upon. In this regard the NRC staff considers the licensees statement to be too vague to be evaluated.
Include information related to the administrative controls and process checks being relied upon to preclude the multiple misloading of fresh fuel in Region 2.
24 The licensee performed calculations for the mislocated, single misload, and the multiple misload accidents. The licensees calculations show the multiple misload accident to be limiting. This is consistent with other SFP NCS analysis.
However, the licensees calculations used depleted 6.5 wt/%
U235 fuel. As previously stated CASMO5 is not approved for above 5.0 wt/% fuel. It is unclear how this affects the determination of the SFP soluble boron requirement. Include information related to the licensees calculations to determine the impact on the SFP soluble boron requirement determination.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION No.
LAR Ref. No.
Document Title 25 The licensee stated, The [MCNP] validation methodology was based on recommendations contained in NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculation Methodology." Using NUREG/CR-6698 is consistent with the guidance in NEI 12-16 R4 and Regulatory Guide 1.240 R0. The licensee did not provide sufficient details regarding the validation analysis. Include information regarding the validation analysis.
26 The description of the NFV lacks specificity and justification.
It is unknown what NFV dimensions were used in the model and how they relate to the actual dimensions. The licensee did not justify using dimensions less than nominal. The licensee did not follow the NEI 12-16 R4 and Regulatory Guide 1.240 R0 with respect to modeling steel below the active fuel and the NFV concrete walls. Include information related to the modeling of the NFV.
27 Compliance with 10 CFR 50.68(b)(2) requires that keff of the NFV not exceed 0.95, at a 95 percent probability, 95 percent confidence level assuming the NFV is flooded with full density water. The licensee determined the keff of its NFV to be 0.9479 if it were flooded with full density water. That is only 0.0021 k of margin to the regulatory limit. As the licensee did not provide sufficient justification for the NRC staff to make an independent evaluation the NRC staff does not have reasonable assurance this amount of margin is accurate. Include information related to the NFV fully flooded analysis.
28 Compliance with 10 CFR 50.68(b)(3) requires that keff of the NFV not exceed 0.98, at a 95 percent probability, 95 percent confidence level assuming the NFV is flooded with an optimum density moderator. The licensee determined the keff of its NFV to be 0.8197 if it were flooded with an optimum density moderator. Thus, the licensee appears to have 0.1603 k of margin to the regulatory limit. As the licensee did not provide sufficient justification for the NRC staff to make an independent evaluation the NRC staff does not have reasonable assurance this amount of margin is accurate. Include information regarding the NFV optimum moderation analysis.
As the licensee adds information to the electronic reading room, it should provide a cross reference to the specific NRC item number listed above. The NRC staff acknowledges and will observe appropriate handling and protection of proprietary information made available for the audit. The NRC staff will not remove non-docketed information from the audit site or web portal.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.0 AUDIT TEAM Joel Wiebe, Project Manager Joel.Wiebe@nrc.gov Kent Wood, SFNB, Technical Reviewer Kent.Wood@nrc.gov Joshua Wilson, STSB, Technical Reviewer Joshua.Wilson@nrc.gov Derek Scully - SCPB, Technical Reviewer Derek.Scully@nrc.gov Thang Thawn - SCPB, Technical Reviewer Thang.Thawn@nrc.gov 6.0 LOGISTICS The audit will be conducted remotely using video and teleconferencing and a secure, online portal, established by the licensee. The audit will begin within 2 weeks of the date of this audit plan and last until the LAR is dispositioned by the NRC. The NRC will establish audit meeting(s)
(e.g., a single, multiday audit meeting; periodic audit meetings throughout the audit period) on mutually agreeable dates and times to discuss information needs and questions arising from the NRCs review of the audited items. The NRCs licensing project manager will inform the licensee of audit meeting dates when they are established, including the date of an audit kickoff meeting.
7.0 DELIVERABLES The NRC staff will issue an audit summary report within 90 days of the audit exit.
ML24113A124 (Package)
ML23362A022 (Proprietary)
ML24113A127 (Non-Proprietary)
OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/SNFB/BC NAME JWiebe SRohrer SKrepel DATE 12/27/2023 1/2/2024 4/11/2024 OFFICE NRR/DSS/SCPB/BC NRR/DSS/STSB/BC(A)
NRR/DORL/LPL3/BC NAME DScully SMehta JWhited DATE 1/3/2024 1/8/2024 4/12/2024 OFFICE NRR/DORL/LPL3/PM NAME JWiebe DATE 4/22/2024