L-2023-077, License Amendment Request 277 Updated Spent Fuel Pool Criticality Analysis
| ML23285A035 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 10/11/2023 |
| From: | Strand D Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-2023-077 | |
| Download: ML23285A035 (42) | |
Text
U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington DC 20555-0001 RE:
Turkey Point Nuclear Plant, Unit 3 and 4 Docket Nos. 50-250 and 50-251 Renewed Facility Operating Licenses DPR-31 and DPR-41 License Amendment Request 277 Updated Spent Fuel Pool Criticality Analysis October 11, 2023 L-2023-077 10 CFR 50.90 Reference
- 1. Westinghouse to NRC, "Transmittal of WCAP-18830-P/NP "Turkey Point Fuel Storage Criticality Analysis for 24 Month Cycles" to Support a License Amendment Request from FPL - License Amendment Request 277 Updated Spent Fuel Pool Criticality Analysis,"
September 22, 2023. (ADAMS Accession No. ML23265A548)
- 2. NRC to FPL, "Turkey Point Nuclear Generating Unit Nos. 3 And 4 -
Issuance of Amendment Nos. 297 and 290 Regarding Conversion to Improved Standard Technical Specifications," September 27, 2023 (ADAMS Accession No. ML23234A192)
Pursuant to 10 CFR Part 50.90, Florida Power & Light Company (FPL) hereby requests amendments to Renewed Facility Operating Licenses (RFOLs) DPR-31 and DPR-41 for Turkey Point Nuclear Plant Units 3 and 4 (Turkey Point), respectively. The proposed license amendments revise the Turkey Point Technical Specifications (TS) by incorporating changes to TS 3.7.13 "Fuel Storage Pool Boron Concentration," TS 3.7.14 "Spent Fuel Storage," and TS 4.3 "Fuel Storage" to allow for an updated spent fuel pool criticality safety analysis which accounts for the impact on the spent fuel from a proposed transition to 24-month fuel cycles.
The enclosure to this letter provides FPL's evaluation of the proposed change. Attachment 1 provides the Turkey Point TS pages marked up to show the proposed changes. Note that the Attachment 1 TS markups are based upon the recently approved TS in Reference 2, which shall be implemented within 180 days of approval (9/27/23). Attachment 2 provides the Turkey Point TS Bases pages marked up to show the proposed changes. The TS Bases markups are provided for information only and will be incorporated in accordance with the Turkey Point TS Bases Control Program upon implementation of the approved license amendments.
In addition, Reference 1 provides the detailed technical basis for the proposed changes to TSs 3. 7.13, 3.7.14, and 4.3. This information was provided under separate docket as explained in the reference transmittal letter.
FPL has determined that the proposed changes do not involve a significant hazards consideration pursuant to 10 CFR 50.92(c), and there are no significant environmental impacts associated with the proposed change. The Turkey Point Onsite Review Group has reviewed the proposed license amendments. In accordance with 10 CFR 50.91 (b)(1 ), a copy of the proposed license amendments is being forwarded to the State designee for the State of Florida.
FPL requests that the proposed change is processed as a normal license amendment request with approval within one year of acceptance to support implementation beginning with the Unit 4 Spring 2025 refuel outage. Once approved, the amendments shall be implemented on a forward fit basis by no later than the Unit 3 spring 2026 reload campaign and the Unit 4 spring 2025 reload campaign, respectively.
Florida Power & Light Company 9760 SW 344th Street, Homestead, FL 33035
Turkey Point Nuclear Plant Docket Nos. 50-250 and 50-251 This letter contains no regulatory commitments.
L-2023-077 Page 2 of 2 Should you have any questions regarding this submission, please contact Mr. Kenneth Mack, Fleet Licensing Manager, at 561-904-3635.
I declare under penalty of perjury that the foregoing is true and correct.
~
Executed on the _ll_ day of October 2023.
V~:57/
Dianne Strand General Manager, Regulatory Affairs
Enclosure:
Enclosure - Evaluation of the Proposed Change Attachments to
Enclosure:
- Redline/strikeout copies of TS 3.7.13, TS 3.7.14, and TS 4.3 - Redline/strikeout copies of TS 3. 7.13 Bases and TS 3. 7.14 Bases cc: USNRC Regional Administrator, Region II USNRC Project Manager, Turkey Point Nuclear Plant USNRC Senior Resident Inspector, Turkey Point Nuclear Plant Mr. Clark Eldredge, Florida Department of Health
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 1 of 14 EVALUATION OF THE PROPOSED CHANGE Turkey Point Nuclear Plant Unit 3 and Unit 4 License Amendment Request 277, Updated Spent Fuel Pool Criticality Analysis 1.0
SUMMARY
DESCRIPTION............................................................................................................. 2 2.0 DETAILED DESCRIPTION............................................................................................................. 2 2.1 System Design and Operation............................................................................................ 2 2.2 Current Technical Specifications Requirements................................................................. 4 2.3 Reason for the Proposed Change...................................................................................... 4 2.4 Description of the Proposed Change.................................................................................. 4
3.0 TECHNICAL EVALUATION
............................................................................................................ 6
4.0 REGULATORY EVALUATION
..................................................................................................... 10 4.1 Applicable Regulatory Requirements/Criteria................................................................... 10 4.2 Precedent.......................................................................................................................... 11 4.3 No Significant Hazards Consideration.............................................................................. 12 4.4 Conclusion........................................................................................................................ 14
5.0 ENVIRONMENTAL CONSIDERATION
........................................................................................ 14
6.0 REFERENCES
............................................................................................................................... 14
- Redline/strikeout copies of TS 3.7.13, TS 3.7.14 and TS 4.3
- Redline/strikeout copies of TS 3.7.13 Bases and TS 3.7.14 Bases
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 2 of 14 1.0
SUMMARY
DESCRIPTION Florida Power & Light Company (FPL) requests amendments to Renewed Facility Operating Licenses (RFOL) DPR-31 and DPR-41 for Turkey Point Nuclear Plant Units 3 and 4 (Turkey Point),
respectively. The proposed license amendments revise the Turkey Point Technical Specifications (TS) by incorporating changes to TS 3.7.13 Fuel Storage Pool Boron Concentration, TS 3.7.14 Spent Fuel Storage, and TS 4.3 Fuel Storage to allow for an updated spent fuel pool criticality safety analysis which accounts for the impact on the spent fuel from a proposed transition to 24-month fuel cycles.
2.0 DETAILED DESCRIPTION 2.1 System Design and Operation Turkey Point Updated Final Safety Analysis Report (UFSAR), Section 9.5.2, describes the spent fuel storage configuration and the associated spent fuel pit (SFP) criticality analyses for the unitized separate spent fuel pools.
The SFP is made up of two fuel storage rack designs (regions), Region I and Region II racks. The SFP storage capacity is 1,535 fuel assemblies (approximately 9 full cores) including 131 spent or fresh fuel assemblies in the cask area rack and miscellaneous fuel handling tools. The spent fuel racks are freestanding and are free to move on the pool liner floor during a seismic event.
Turkey Point has an absorber surveillance program for the Metamic inserts used in Region II as well as for the Boral panels in the Cask Area Racks. The program is detailed in UFSAR Section 17.2.2.27, delineating both the testing requirements and frequency. The Metamic inserts program relies on coupons as well as visual inspections and testing of used inserts, whereas the Boral program relies on in-situ testing.
Region I Spent Fuel Pit Racks The SFP Region I racks consist of two 8 x 11 modules and one 10 x 11 module for a total of 286 storage locations. Region I is the high-enrichment, core off-load region.
Region I racks in the spent fuel pit permit storage of fresh and irradiated fuel assemblies.
The Region I storage racks are free-standing, seismically qualified components composed of individual storage cells made of stainless steel. These racks have a neutron absorbing material, Boraflex, which is attached to each cell. No credit is taken for Boraflex in the criticality safety analysis. The cells within a module are interconnected by grid assemblies to form an integral structure. The modules are neither anchored to the floor nor braced to the pool walls. The grid assemblies maintain the 10.6-inch centerline-to-centerline spacing between cells [TS 4.3.1 critical design feature].
As described in Section 9.5.2.2 of the UFSAR, without credit for Boraflex as a neutron absorber, the placement of fuel into Region I is controlled based on specific loading patterns, defined by four allowable 2 x 2 arrays. The use of Rod Cluster Control Assemblies (RCCAs) as a neutron absorber is credited in two of the defined loading patterns.
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 3 of 14 Region II Spent Fuel Pit Racks The SFP Region II racks consist of three 9 x 13 modules, one modified 9 x 13 module, one 9 x 14 module (minus 25 cells removed from fuel storage service due to their proximity to the cooling system return piping discharge flow path), three 10 x 13 modules, and one 10 x 14 module for a total storage capacity of 1093 fuel assemblies.
The Region II storage racks are the same basic design as the SFP Region I racks with the following exceptions:
a) The cells are assembled in a checkerboard pattern with a 9.0-inch centerline-to-centerline spacing [TS 4.3.1 critical design feature].
b) The cells are welded to the base support assembly and to one another to form an integral structure without use of grids as used in SFP Region I racks.
c) The Region II racks were manufactured with Boraflex poison panels installed; however, due to the degradation, Boraflex poison panels are no longer credited in the criticality analysis per Section 9.5.2.2 of the UFSAR. As a consequence, criticality control within the Region II spent fuel racks is provided by following prescribed loading patterns which include specific fuel categories based upon enrichment, burnup and cooling time, along with the use of a combination of RCCAs, empty cells and Metamic inserts. Metamic inserts are a metal matrix composite of aluminum and boron carbide. The Metamic inserts are manufactured with a nominal boron carbide content of 0.0160 g/cm2 and a minimum of 0.0150 g/cm2. Metamic inserts will be installed in Region II racks cells between the fuel assembly and the inside cell wall, when required by the specific loading pattern.
Cask Area Rack The cask area of the spent fuel pit is designed for the installation of a fuel transfer cask to allow fuel transfer operations. However, to provide increased fuel storage capability, a rack may be installed in the cask area when not performing fuel transfer operations. The Region I cask area rack is an 11 x 12 module with 131 storage locations (one location has been omitted to allow the placement of a fuel handling tool). The cask area rack is of the same basic design as the spent fuel pit Region I racks with the following exceptions:
a) The center-to-center spacing for the cells is 10.1 E-W and 10.7 N-S [TS 4.3.1 critical design feature]
b) Boral panels are installed as a neutron absorber instead of Boraflex.
c) Bearing pads, 12 square, are installed between the rack leveling screws and the pit floor to provide sliding contact and distribute the rack weight.
When fuel transfer operations are necessary, fuel assemblies stored in the cask area rack are relocated to the spent fuel pit racks and the cask area rack is removed, decontaminated, as necessary, and placed in storage.
The design of the spent fuel pit racks incorporates the requirements of, and are in accordance with, USNRC "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 14, 1978, as amended by the NRC letter dated January
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 4 of 14 18, 1979, and the applicable portions of the NRC Regulatory Guides, Standard Review Plan Sections, and published standards.
The design criteria for preventing criticality in the SFP storage racks and the cask area rack are based on the 95/95 rule (10 CFR 50.68(b)). The criteria are as follows:
- 2. keff less than or equal to 0.95 with the presence of a defined level of soluble boron in the spent fuel pit water.
2.2 Current Technical Specifications Requirements FPL proposes changes to the spent fuel storage that affects TS 3.7.13 Fuel Storage Pool Boron Concentration, TS 3.7.14 Spent Fuel Storage, and TS 4.3 Fuel Storage for Turkey Point Units 3 and 4.
2.3 Reason for the Proposed Change The purpose of the proposed changes to TS 3.7.13, TS 3.7.14, and TS 4.3 is to update the SFP criticality analysis. The updated analysis follows the guidance in NEI 12-16 (Reference 6.1) as endorsed in RG-1.240, Fresh and Spent Fuel Pool Criticality Analyses.
(Reference 6.2). The proposed changes account for the impact on the spent fuel from a proposed transition to 24-month cycles.
2.4 Description of the Proposed Change TS 3.7.13, Fuel Storage Pool Boron Concentration The proposed changes revise TS 3.7.13 based on the updated criticality analysis, Turkey Point Fuel Storage Criticality Analysis for 24 Month Cycles (Reference 6.5, - WCAP-18830-P, Proprietary Version and Attachment 3 - WCAP-18830-NP, Non-Proprietary Version), and the resulting update to the boron concentration requirements.
The proposed changes to TS 3.7.13 include:
A revision to the water boration level from 2300 ppm to 2350 ppm for the limiting condition of operation (LCO) 3.7.13 that requires the fuel storage pool boron concentration to be maintained above the specified amount.
Note: Reference 6.5, Attachments 2 and 3 are applicable to all references to the updated criticality safety analysis from hereon.
TS 3.7.14, Spent Fuel Storage The proposed changes revise TS 3.7.14 based on the updated criticality analysis. This includes the addition of a new fuel storage configuration (array) as well as updated burnup limits and other storage restrictions for all fuel storage configurations.
The proposed changes to TS 3.7.14 include:
Replacing Tables 3.7.14-1 and 3.7.14-2 with new Tables 3.7.14-1 through 3.7.14-3 that show the coefficients to calculate the minimum required fuel assembly burnup as a function of enrichment and decay time for pre-EPU non-blanketed
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 5 of 14 assemblies, mid-enriched blanketed assemblies, and non-blanketed assemblies for each of the different fuel categories (I-3, I-4, II-1, II-2, II-3, II-4, II-5 and II-6).
Updating the LCO references to Tables 3.7.14-1 through -3 throughout the text accordingly.
Renaming the current Table 3.7.14-3 to become Table 3.7.14-4 and updating the LCO references to the table throughout the text.
Adding a row at the bottom of the Region II section of new Table 3.7.14-4, placing II-6 in the middle column, and shifting Low Reactivity down to align with the new II-6 row.
Deleting the current Table 3.7.14-4 that shows the IFBA requirements for fuel category I-2.
The addition of a 6th requirement to LCO 3.7.14.b for fresh and irradiated fuel assemblies in the Region I or Region II racks that states Fuel in Category I-2 shall meet the minimum IFBA requirement given by the following equation: Minimum IFBA = -22.222*En2 + 272.22*En - 711.96, where En is equal to the fresh I-2 enrichment and greater than 3.78 weight percent U-235.
Replacing any references throughout the text and in Figures 3.7.14-1 through 3.7.14-3 to the current Table 3.7.14-4 with a reference to the new LCO 3.7.14.b.6.
Adding a definition and illustration to Figure 3.7.14-2 for the newly proposed Array II-E.
Fixing some editorial errors in the definitions for Array II-B, Array II-C, and Array II-D within Figure 3.7.14-2.
Replacing current Figure 3.7.14-3 with new Figure 3.7.14-3.
TS 4.3, Fuel Storage The proposed change to TS 4.3 revises the section based on the updated criticality safety analysis and the resulting update to the boron concentration requirements.
The proposed changes to TS 4.3 include:
A revision to the water boration level from 500 ppm to 550 ppm for the requirement that states keff must be less than or equal to 0.95 if the spent fuel storage racks are fully flooded with water borated to the specified amount, which includes an allowance for biases and uncertainties as described in Section 9.5 of the UFSAR.
Redline/strikeout copies of TS 3.7.13, TS 3.7.14, and TS 4.3 are included in Attachment
- 1. Redline/strikeout copies of TS 3.7.13 Bases and TS 3.7.14 Bases, are provided (for information only) in Attachment 2.
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 6 of 14
3.0 TECHNICAL EVALUATION
This License Amendment Request (LAR) was modeled based on the guidance of NEI 12-16 (Reference 6.1). FPL used the NEI guidance to ensure that the proper considerations were made in the analysis and controls. The updated criticality safety analysis also reflects the guidance of RG 1.240 (Reference 6.2), which endorses, with clarifications and exceptions, NEI 12-16, Revision 4.
This section includes a brief statement related to each applicable topic discussed in NEI 12-16 (Reference 6.1) and summarizes the analysis or proposed controls applicable to each area. Also included is a discussion on the proposed TS changes. Supporting details are found in Reference 6.5, Attachments 2 and 3, as referenced below.
Acceptance Criteria NEI 12-16, Section 2 (Reference 6.1) describes the NRC acceptance criteria for spent fuel pool storage of new and use fuel for pools where credit for soluble boron is taken as follows:
The criticality safety analyses must meet two independent limits:
- a. With the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water, the keff must remain below 1.0 (subcritical),
at a 95 percent probability, 95 percent confidence level.
- b. With the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with borated water, the keff must not exceed 0.95, at a 95 percent probability, 95 percent confidence level.
Section 2.1 of the updated criticality analysis (Reference 6.5) provides the same acceptance criteria for the updated Turkey Point SFP criticality analysis.
Computer Codes NEI 12-16, Section 3 (Reference 6.1) describes the different types of computer codes that may be used in a criticality analysis. This section also discusses the validation of the computer codes used in the criticality analysis. The licensee needs to state which codes were utilized along with the type/version of cross-section libraries.
The updated criticality safety analysis utilized in this submittal employs the following computer codes and cross-section libraries:
- 1. The two-dimensional (2-D) transport lattice code PARAGON Version 1.4.3, as documented in WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON (Reference 6.3) and its cross-section library based on Evaluated Nuclear Data File Version VI.3 (ENDF/B-VI.3)
- 2. Scale Version 6.2.4 with the ENDF/B-VII.1 252 group cross-section library.
The Computer Codes used in this application are discussed in Section 2.3 of the updated criticality safety analysis (Reference 6.5). PARAGON is generically approved for depletion calculations (Reference 6.3). The applicable specific validation of Scale Version 6.2.4 is provided in Appendix A of the updated criticality analysis (Reference 6.5).
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 7 of 14 Reactivity Effects of Depletion NEI 12-16, Section 4 (Reference 6.1) describes appropriate considerations for calculating reactivity effects of fuel depletion. Significant parameters that could impact reactivity of used fuel in depletion analyses are:
Power, moderator temperature, and fuel temperature during depletion Soluble boron during depletion Presence of burnable absorbers Rodded operation Cooling time and other depletion parameters are also discussed in NEI 12-16, Section 4 (Reference 6.1).
The depletion analysis is described in Section 4.1 of the updated criticality safety analysis and describes the methods used to determine conservative and bounding inputs for the generation of isotopic number densities. Controls which ensure future fuel designs satisfy the assumptions of the analysis are discussed in the Licensee Controls section, below. Depletion uncertainty is discussed in Section 4.2.5.1.5 of the updated criticality safety analysis.
Fuel Assembly and Storage Rack Modeling NEI 12-16, Section 5 (Reference 6.1) describes generally acceptable methods of modeling fuel assemblies and fuel storage racks, including considerations for rack neutron absorbers.
All fuel assemblies operated or planned for operation at Turkey Point are considered in the updated criticality safety analysis. Details of the fuel assembly designs are provided in Section 3.2 and 3.3 of the updated criticality analysis (Reference 6.5). The updated criticality safety analysis allows fuel assemblies which have not operated in the reactor to take credit for the presence of zirconium diboride in the integral fuel burnable absorber (IFBA).
The spent fuel pool is made up of two fuel storage rack designs (regions), Region I and Region II racks. The SFP storage capacity is 1,535 fuel assemblies (approximately 9 full cores) including 131 spent or fresh fuel assemblies in the cask area rack and miscellaneous fuel handling tools. The storage racks have a neutron absorbing material, Boraflex, which is attached to each cell. No credit is taken for the Boraflex in the criticality analysis. However, credit is taken for the presence of soluble boron in the SFP. Details of the storage rack parameters are provided in Sections 3.4.2 and 3.4.3 of the updated criticality safety analysis (Reference 6.5).
The fuel and storage rack manufacturing tolerances, eccentric fuel assembly positioning bias, and SFP temperature bias are included in the updated criticality safety analysis through either analysis or use of bounding values. Details of these items are provided in Section 4.2.5 of the updated criticality safety analysis (Reference 6.5).
The axial burnup distribution and reactor record burnup uncertainties are considered in the updated criticality safety analysis. Details of the axial burnup distribution are provided in Section 4.1.2.2.4 of Reference 6.5, Attachments 2 and 3. The details for the burnup uncertainties are provided in Section 4.2.5.1.4 of Reference 6.5, Attachments 2 and 3.
The New Fuel Storage Rack analysis is described in Section 5.9 of the updated criticality safety analysis (Reference 6.5, Attachments 2 and 3). The rack analysis considers both fully flooded and optimal moderation conditions. The New Fuel Storage Rack Requirements are as follows:
No restrictions up to and including 4.50 wt. % 235U.
16 IFBA rods required for fuel greater than 4.50 wt. % 235U (up to 5.0 wt. % 235U).
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 8 of 14 Configuration Modeling and Soluble Boron Credit NEI 12-16, Section 6 (Reference 6.1) describes considerations for configuration modeling, including a description of normal conditions, interface considerations, and abnormal / accident conditions which should be considered. NEI 12-16, Section 7 (Reference 6.1) describes considerations for soluble boron credit under normal and accident conditions, and considerations for a boron dilution accident.
The updated criticality safety analysis demonstrates acceptable results keff for both normal conditions (Section 5.8.1) and accident conditions (Section 5.8.2). Normal conditions include normal storage, fuel movement, and other procedurally controlled activities in the SFPs. Fuel assembly storage arrays, which define allowable storage, are defined in Section 4.2.1 of the updated criticality safety analysis. The addition of a new fuel storage array to the currently approved fuel storage arrays has been incorporated into the proposed TS changes. Controls which ensure that the proposed TS limitations for storage are maintained are discussed in the Licensee Controls section below.
Interface considerations are described in Section 4.2.7 of the updated criticality safety analysis.
The interfaces are the locations where there is a change in either the storage racks or the storage requirements of the fuel in question. Only the intra-region interfaces are evaluated because all racks are of the same design and no pool region interfaces are present. Each storage cell can be part of four different storage arrays. Compliance with the storage arrays in the TS will ensure acceptable boundary cells at the interface.
Every 2x2 configuration matches an analyzed condition, and therefore no interface-specific analyses are required. Gaps between the same region rack modules are conservatively neglected, i.e, cells located across a rack-to-rack gap are considered the same as cells directly facing each other within a rack. The configuration where Region II cells face Region I rack modules require additional analyses and are discussed in Section 4.2.7.2 of the updated criticality safety analysis.
No special considerations need to be given to cells facing the pool walls or other racks.
Soluble boron is credited in the Turkey Point Units 3&4 SFPs to keep keff < 0.95 under all normal and credible accident scenarios. Accident conditions considered are the following:
Misloaded fresh fuel assembly or assemblies in a storage rack Inadvertent removal of an absorber insert Spent fuel pool temperature greater than normal operating range Loss of water gap between Region I and Region II due to seismic event Dropped fresh fuel assembly Misplaced fuel assembly Misloaded cask area storage rack Misloaded upender Per Section 9.5.2.3 of the Turkey Point UFSAR, most postulated accidents in the spent-fuel rack will not result in an increase in reactivity. All of the accident conditions listed below the misloaded fresh fuel assembly or assemblies in a storage rack in the list above have been determined to be less limiting than the fresh fuel assembly misload case.
Single misload events were previously analyzed; however, the updated criticality safety analysis also includes analysis of a multiple misload accident scenario in accordance with NEI 12-16, Section 6.3.5. The inclusion of this analysis does not imply the creation of the possibility of a new accident but expands the boundaries of the analyzed accident conditions to ensure that all potential accidents are properly considered. Past industry experience has shown the potential for multiple misload accidents to varying degrees. The analysis of a full pool multiple misload accident bounds all credible accident scenarios.
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 9 of 14 The multiple misload fuel assembly accident is the limiting analyzed accident condition. This event involves misloading multiple fuel assemblies in series due to a common cause.
As a result of this more limiting event now being analyzed, the boron concentration required to maintain keff 0.95 in TS 3.7.13 Fuel Storage Pool Boron Concentration has been increased from 2300 ppm to 2350 ppm. This updated concentration includes the NEI 12-16 Rev. 4 suggested 50 ppm of soluble boron margin (in addition to rounding). Note that in the TS Bases, the discussion regarding significant margin between the calculated ppm requirement (1700 ppm) and the current TS value (2300 ppm) has been removed as a result of analyzing the multiple misload event in its most bounding configuration.
For the limiting normal condition, 550 ppm of soluble boron is credited to ensure the maximum keff satisfies the acceptance criteria of keff 0.95. The increase from 500 ppm to 550 ppm of soluble boron in TS 4.3 Fuel Storage is the result of the updated criticality safety analysis as required for the new fuel management strategy. This updated concentration includes the NEI 12-16 Rev. 4 suggested 50 ppm of soluble boron margin (in addition to rounding).
A spent-fuel pool dilution event has been previously evaluated by FPL and was determined not to be a credible event for Turkey Point (refer to Section 9.5.2.3 of the UFSAR).
Calculation of Maximum keff NEI 12-16, Section 8 (Reference 6.1) describes that the maximum keff is determined by adding to the nominal calculated keff any biases that may exist in the methodology and the applicable uncertainties using the formula described below, for comparison to the acceptance limits.
kk BiasUncertainty
The updated criticality safety analysis demonstrates that the keff, including all applicable biases and uncertainties which account for the statistical 95/95 confidence levels, satisfy the acceptance criteria. The sum of biases is additive while the sum of uncertainties are statistically added as the root sum square of the individual reactivity uncertainties as described in Section 4.2.5 of the updated criticality safety analysis.
Licensee Controls NEI 12-16, Section 9 (Reference 6.1) describes controls intended to ensure that the conditions evaluated in the nuclear criticality safety analysis are and remain bounding to the current plant operating parameters. It discusses procedural controls for fuel storage and for planning and performance of fuel movements, new (future) fuel types, and pre-and post-irradiation fuel characterization.
In conjunction with the implementation of the proposed license amendments, the reactivity management procedural controls are changed to be consistent with the updated criticality safety analysis. These controls ensure the spent fuel pool configuration and other applicable conditions evaluated in the updated criticality safety analysis remain bounding when compared to current fuel design and plant operating parameters. Specifically, these procedural controls ensure:
- 1. TS 3.7.13, 3.7.14, and TS 4.3 compliance is maintained whenever any fuel assembly is stored in the spent fuel pit. Controls are established to ensure that all fuel movement plans into the spent fuel pool are prepared in a manner which ensures continual compliance with the
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 10 of 14 limitations of proposed TS 3.7.13, TS 3.7.14 and TS 4.3, including all intermediate steps during fuel movement.
- 2. A misloading event beyond the analyzed accident conditions is not credible. Controls are established to ensure that an error in the fuel move planning does not have the potential to result in a misloading accident which is not bounded by the updated criticality safety analysis.
- 3. Assumptions related to fuel characterization and reactor operation remain valid. Controls are established to ensure that conditions evaluated in the updated criticality safety analysis will remain bounding for both future fuel design changes (pre-irradiation fuel characterization) and future operating conditions (post-irradiation fuel characterization).
Conclusion In conclusion, the proposed changes to TS 3.7.13, TS 3.7.14, and TS 4.3 allow safe storage of spent fuel at Turkey Point.
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria (Note: The General Design Criterion (GDC) used during the licensing of Turkey Point were based on the 1967 Atomic Energy Commission Proposed General Design Criterion (1967 Proposed GDC) and predate 10 CFR Part 50, Appendix A.)
1967 Proposed General Design Criterion (GDC) 68 - Adequate shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities.
1967 Proposed GDC 69 - Provisions shall be made in the design of fuel and waste storage facilities such that no undue risk to the health and safety of the public could result from an accidental release of radioactivity.
1967 Proposed GDC 66 - Criticality in the new and spent fuel storage pits shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls.
Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commission's regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36. The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety system settings, and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.
The requirements for system operability during movement of irradiated fuel are included in the TSs in accordance with 10 CFR 50.36(c)(2), Limiting Conditions for Operation. As required by 10 CFR 50.36(c)(4), design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of 10 CFR 50.36.
This amendment request concerns 10 CFR 50.36(c)(2) and 10 CFR 50.36(c)(4).
At a regulatory level, 10 CFR 50.68(a) requires licensees to select one of two options to satisfy criticality accident requirements: (1) 10 CFR 70.24, or (2) 10 CFR 50.68(b)
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 11 of 14 as highlighted in RIS 2005-05, Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations, (Reference 6.4).
As described in UFSAR Subsections 9.5.1.1 and 9.5.2.1, Turkey Point has elected to comply with the requirements of 10 CFR 50.68(b), which includes restrictions on the reactivity of stored fresh (new) fuel and on the reactivity of stored spent fuel.
Additional guidance is available in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," particularly Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling," Revision 3, issued March 2007. Section 9.1.1 provides the existing recommendations for performing the review of the nuclear criticality safety analysis of SFPs.
NRC guidance for performing criticality analysis is contained in Regulatory Guide (RG) 1.240, Fresh and Spent Fuel Pool Criticality Analyses (Reference 6.2). This guidance endorses NEI-12-16, Rev. 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants (Reference 6.1).
4.2 Precedent License amendment requests and applicable requests for additional information (RAIs) associated with the following NRC Issuance of Amendment letters have been used in the development of the criticality safety analysis and the appropriate sections of the amendment request:
Comanche Peak Units 1 and 2 License Amendment Request for Spent Fuel Pool Technical Specification Changes in a letter dated July 1, 2014, Comanche Peak Nuclear Power Plant, Units 1 and 2 - Issuance of Amendments RE: Revision to Technical Specifications 3.7.16, Fuel Storage Pool Boron Concentration, 3.7.17, Spent Fuel Assembly Storage, 4.3, Fuel Storage, and 5.5, Programs and Manuals (TAC NOS. MF1365 and MF1366), (ADAMS Accession No. ML14160A035).
Palo Verde Units 1, 2, and 3 License Amendment Request for Spent Fuel Pool Technical Specification Changes in a letter dated July 28, 2017, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments to Revise Technical Specifications to Incorporate Updated Criticality Safety Analysis (CAC NOS.
MF7138, MF7139, and MF7140), (ADAMS Accession No. ML17188A412).
Prairie Island Units 1 and 2 License Amendment Request for Spent Fuel Pool Criticality Technical Specification Changes in a letter dated November 30, 2017, Prairie Island Nuclear Generating Plant, Units 1 and 2 - Issuance of Amendment Revising Spent Fuel Pool Criticality Technical Specification (CAC NOS. MF7121 and MF7122, EPID L-2015-LLA-0002), (ADAMS Accession No. ML17334A178).
Joseph M. Farley Units 1 and 2 License Amendment Request for Spent Fuel Pool Criticality Technical Specification Changes in a letter dated October 6, 2020, Joseph M. Farley Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos. 229 and 226 Regarding Spent Fuel Pool Criticality Safety Analysis (EPID L-2019-LLA-0212),
(ADAMS Accession No. ML20196L929).
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 12 of 14 4.3 No Significant Hazards Consideration Florida Power & Light Company (FPL) requests amendments to Renewed Facility Operating Licenses (RFOL) DPR-31 and DPR-41 for Turkey Point Nuclear Plant Units 3 and 4 (Turkey Point), respectively. The proposed license amendments revise the Turkey Point Technical Specifications (TS) by incorporating changes to TS 3.7.13 Fuel Storage Pool Boron Concentration, TS 3.7.14 Spent Fuel Storage, and TS 4.3 Fuel Storage to allow for an updated spent fuel pool criticality safety analysis which accounts for the impact on the spent fuel from a proposed transition to 24-month fuel cycles.
As required by 10 CFR 50.91(a), FPL has evaluated the proposed change using the criteria in 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration. An analysis of the issue of no significant hazards consideration is presented below:
(1)
Do the proposed amendments involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed amendment was evaluated for impact to the single misload events that were previously analyzed and determined to be the most limiting for Turkey Point; however, the updated criticality safety analysis also includes analysis of a multiple misload accident scenario in accordance with NEI 12-16, Section 6.3.5.
The inclusion of this analysis does not imply the creation of the possibility of a new accident but expands the boundaries of the analyzed accident conditions to ensure that all potential accidents are properly considered. The analysis of a full pool multiple misload accident bounds all credible accident scenarios.
Operation in accordance with the proposed amendment will not change the probability of a loss of spent fuel pool cooling because the changes in the criticality safety analysis have no bearing on the systems, structures, and components involved in initiating such an event. A criticality safety analysis of the limiting fuel loading configuration confirmed that the condition would remain subcritical for a range of normal and accident conditions. The effects of the accident conditions are bounded by the multiple fuel assembly misload accident.
Operation in accordance with the proposed amendment will not change the probability of a fuel assembly being dropped into an already loaded storage cell because fuel movement will continue to be controlled by approved fuel handling procedures. The consequences of a dropped fuel assembly are not changed; there will continue to be significant separation between the dropped fuel assembly and the active regions of the fuel assemblies. The effects of this accident are bounded by the multiple fuel assembly misload accident.
Operation in accordance with the proposed amendment will not change the probability of a fuel assembly misloading because fuel movement will continue to be controlled by approved fuel selection and fuel handling procedures. These procedures continue to require identification of the initial and target locations for each fuel assembly and fuel assembly insert that is moved. The consequences of a fuel misloading event are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for the multiple fuel assembly misload accident.
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 13 of 14 Operation in accordance with the proposed amendment will not change to the probability of a fuel assembly being incorrectly categorized because all calculations including categorization will continue to be controlled by approved process requiring preparation and verification. Although one additional criticality fuel design was added, this only applies to the legacy assemblies and those categorizations will be performed and verified once and not changed again. All new assemblies will have the same category (criticality fuel design). The effects of this accident are bounded by the multiple fuel assembly misload accident.
Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
(2)
Do the proposed amendments create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The potential for criticality in the spent fuel pool is not a new or different type of accident. Storage configurations allowed by Technical Specifications 3.7.13, 3.7.14, and 4.3 have been analyzed with appropriate codes and methods to demonstrate that the pool remains subcritical.
The new criticality safety analysis includes analysis of a multiple misload accident scenario; only single misload events were previously analyzed. The inclusion of this analysis does not imply the creation of the possibility of a new accident, but simply expands the boundaries of the analyzed accident conditions to ensure that all potential accidents are properly considered.
There is no significant change in plant configuration, equipment design or usage of plant equipment. The updated criticality safety analysis assures that the pool will continue to remain subcritical.
Therefore, the proposed amendments do not create the possibility of a new or different kind of accident from any previously evaluated.
(3)
Do the proposed amendments involve a significant reduction in a margin of safety?
Response: No The proposed change was evaluated for its effect on current margins of safety as they relate to criticality, structural integrity, and spent fuel heat removal capability.
The margin of safety for subcriticality required by 10 CFR 50.68 is unchanged.
New criticality analysis confirms that operation in accordance with the proposed amendment continues to meet the required subcriticality margins. The proposed change does not affect spent fuel heat generation or the spent fuel pool cooling systems. Therefore, the proposed changes do not involve a significant reduction in the margin of safety.
Therefore, the proposed amendments do not involve a significant reduction in a margin of safety.
Based upon the above analysis, FPL concludes that the proposed license amendment does not involve a significant hazards consideration, under the standards set forth in 10 CFR 50.92, Issuance of Amendment, and accordingly, a finding of no significant hazards consideration is justified.
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Page 14 of 14 4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
5.0 ENVIRONMENTAL CONSIDERATION
The proposed amendments revise the Turkey Point Technical Specifications (TS) by updating the SFP criticality analysis. FPL has evaluated the proposed amendments for environmental considerations and determined that the proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendments do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.
6.0 REFERENCES
6.1 NEI-12-16, Rev. 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, September 2019 (ADAMS Accession No. ML19269E069).
6.2 Regulatory Guide (RG) 1.240, Revision 0, Fresh and Spent Fuel Pool Criticality Analyses, March 2021 (ADAMS Accession No. ML20356A127).
6.3 Westinghouse Document WCAP-16045-P-A, Rev. 0, Qualification of the Two-Dimensional Transport Code PARAGON, August 2004 (ADAMS Accession No. ML040780402).
6.4 Regulatory issues Summary 2005-05, Regulatory Issues Regarding Criticality Analyses for Spent Fuel Pools and Independent Spent Fuel Storage Installations, March 23, 2005 (ADAMS Accession No. ML043500532).
6.5 Westinghouse to NRC, Transmittal of WCAP-18830-P/NP Turkey Point Fuel Storage Criticality Analysis for 24 Month Cycles to Support a License Amendment Request from FPL - License Amendment Request 277 Updated Spent Fuel Pool Criticality Analysis, September 22, 2023. (ADAMS Accession No. ML23265A548)
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 Enclosure 1 ATTACHMENT 1 Turkey Point Technical Specifications Page Markups (18 pages follow)
Fuel Storage Pool Boron Concentration 3.7.13 Turkey Point Unit 3 and Unit 4 3.7.13-1 Amendment Nos. 297 and 290 3.7 PLANT SYSTEMS 3.7.13 Fuel Storage Pool Boron Concentration LCO 3.7.13 7KHIXHOVWRUDJHSRROERURQFRQFHQWUDWLRQVKDOOEH 2300 ppm.
APPLICABILITY:
When fuel assemblies are stored in the fuel storage pool and a fuel storage pool verification has not been performed since the last movement of fuel assemblies in the fuel storage pool.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool boron concentration not within limit.
NOTE-------------------
LCO 3.0.3 is not applicable.
A.1 Suspend movement of fuel assemblies in the fuel storage pool.
AND A.2.1 Initiate action to restore fuel storage pool boron concentration to within limit.
OR A.2.2 Initiate action to perform a fuel storage pool verification.
Immediately Immediately Immediately 2350
Fuel Storage Pool Boron Concentration 3.7.13 Turkey Point Unit 3 and Unit 4 3.7.13-2 Amendment Nos. 297 and 290 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify the fuel storage pool boron concentration is within limit.
In accordance with the Surveillance Frequency Control Program This page is for information only. No changes are proposed for this page.
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-1 Amendment Nos. 297 and 290 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Storage LCO 3.7.14 The combination of initial enrichment, burnup, and cooling time of each fuel assembly stored in the spent fuel pit shall be in accordance with the following:
a.
No restrictions on storage of fresh or irradiated fuel assemblies in the cask area storage rack are applicable.
b.
Fuel assemblies stored in Region I and II shall be stored in accordance with the requirements of Figures 3.7.14-1 through 3.7.14-3 with credit for burnup and cooling time taken in determining acceptable placement locations for spent fuel in the two-region spent fuel racks. Fresh and irradiated fuel assemblies in the Region I or Region II racks shall be stored in compliance with the following:
1.
any 2x2 array of Region I storage cells containing fuel shall comply with the storage patterns in Figure 3.7.14-1 and the requirements of Tables 3.7.14-1 and 3.7.14-2, as applicable. The reactivity rank of fuel assemblies in the 2x2 array (rank determined using Table 3.7.14-3) shall be equal to or less reactive than that shown for the 2x2 array.
2.
any 2x2 array of Region II storage cells containing fuel shall:
i.
comply with the storage patterns in Figure 3.7.14-2 and the requirements of Tables 3.7.14-1 and 3.7.14-2, as applicable. The reactivity rank of fuel assemblies in the 2x2 array (rank determined using Table 3.7.14-3) shall be equal to or less reactive than that shown for the 2x2
- array, ii.
have the same directional orientation for Metamic inserts in a contiguous group of 2x2 arrays where Metamic inserts are required, and iii. comply with the requirements of LCO 3.7.14.b.3. for cells adjacent to Region I racks.
3.
Any 2x2 array of Region II storage cells that interface with Region I storage cells shall comply with the rules of Figure 3.7.14-3.
4.
Any fuel assembly may be replaced with a fuel rod storage basket or non-fuel hardware.
3.7.14-1 through 3.7.14-3 3.7.14-4 3.7.14-1 through 3.7.14-3 3.7.14-4
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-2 Amendment Nos. 297 and 290 LCO 3.7.14 (continued) 5.
Storage of Metamic inserts or rod cluster control assemblies (RCCAs) is acceptable in locations designated as empty (water-filled) cells.
APPLICABILITY:
Whenever any fuel assembly is stored in the spent fuel pit.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO not met.
A.1
NOTE--------------
LCO 3.0.3 is not applicable.
Initiate action to move the noncomplying fuel assembly to an acceptable location.
Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify by administrative means the initial enrichment, burnup, and cooling time of the fuel assembly is in accordance with the Figure 3.7.14-1 through Figure 3.7.14-3.
Prior to storing the fuel assembly in Region I or II
- 6. Fuel in Category I-2 shall meet the minimum IFBA requirement given by the following equation:
Minimum IFBA = -22.222*En2 + 272.22*En - 711.96 where En is equal to the fresh I-2 enrichment and greater than 3.78 weight percent U-235.
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-3 Amendment Nos. 297 and 290 Table 3.7.14-1 (page 1 of 1)
Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct)
See Notes 1-4 for use of Table 3.7.14-1 Coeff.
Fuel Category I-3 I-4 II-1 II-2 II-3 II-4 II-5 A1 5.66439153
-14.7363682
-7.74060457
-7.63345029 24.4656526 8.5452608 26.2860949 A2
-7.22610116 11.0284547 5.13978237 10.7798957
-20.3141124
-4.47257395
-18.0738662 A3 2.98646188
-1.80672781
-0.360186309
-2.81231555 6.53101471 2.09078914 5.8330891 A4
-0.287945644 0.119516492 0.0021681285 0.29284474
-0.581826027
-0.188280562
-0.517434342 A5
-0.558098618 0.0620559676
-0.0304713673 0.0795058096
-0.16567492 0.157548739
-0.0614152031 A6 0.476169245 0.0236575787 0.098844889
-0.0676341983 0.243843226
-0.0593584027 0.134626308 A7
-0.117591963
-0.0088144551
-0.0277584786 0.0335130877
-0.0712130368 0.0154678626
-0.0383060399 A8 0.0095165354 0.0008957348 0.0024057185
-0.0040803875 0.0063998706
-0.0014068318 0.0033419846 A9
-47.1782783
-20.2890089
-21.424984 14.6716317
-41.1150
-0.881964768
-12.1780 A10 33.4270029 14.7485847 16.255208
-10.0312224 43.9149156 9.69128392 23.6179517 A11
-6.11257501
-1.22889103
-1.77941882 5.62580894
-9.6599923
-0.18740168
-4.10815592 A12 0.490064351 0.0807808548 0.127321203
-0.539361868 0.836931842 0.0123398618 0.363908736 Notes:
1.
All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the minimum burnup (GWd/MTU) given by the curve fit for the assembly cooling time and initial enrichment. The specific minimum burnup required for each fuel assembly is calculated from the following equation:
Bu = (A1 + A2*En + A3*En2 + A4*En3)* exp [ - (A5 + A6*En + A7*En2 + A8*En3)*Ct ] + A9 + A10*En + A11*En2 + A12*En3 2.
Initial enrichment, En, is the nominal central zone U-235 enrichment. Axial blanket material is not considered when determining enrichment. Any enrichment between 2.0 and 5.0 may be used.
3.
Cooling time, Ct, is in years. Any cooling time between 0 years and 25 years may be used. An assembly with a cooling time greater than 25 years must use 25 years.
4.
This Table applies for any blanketed fuel assembly.
Delete Table 3.7.14-1 and insert revised Tables 3.7.14-1 through 3.7.14-3
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-4 Amendment Nos. 297 and 290 Table 3.7.14-2 (page 1 of 1)
Non-Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling Time (Ct)
See Notes 1-4 for use of Table 3.7.14-2 Coeff.
Fuel Category I-3 I-4 II-1 II-2 II-3 II-4 II-5 A1 2.04088171
-27.6637884
-11.2686777 20.7284208 29.8862876
-83.5409405 35.5058622 A2
-4.83684164 26.1997193 2.0659501 11.9673275
-37.0771132 94.7973724
-30.1986997 A3 2.59801889
-7.2982252 2.66204924
-14.4072388 16.3986049
-31.9583373 11.0102438 A4
-0.300597247 0.723731768
-0.513334362 2.83623963
-2.1571669 3.55898487
-1.27269125 A5
-0.610041808 0.401332891
-0.0987986108
-1.49118695 1.02330848 0.299948492 1.34723758 A6 0.640497159
-0.418616707
-0.0724198633 1.75361041
-1.21889631
-0.312341996
-1.19871392 A7
-0.219000712 0.144304039 0.106248806
-0.659046438 0.467440882 0.107463895 0.352920811 A8 0.0252870451 -0.0154239536
-0.0197359109 0.080884618
-0.0560129443
-0.0108814287
-0.0325155213 A9
-4.48207836
-5.54507376
-1.34620551
-245.825283 12.1549 39.4975573
-5.2576 A10
-2.12118634
-5.76555416
-10.1728821 243.59979
-22.7755385
-50.5818253 10.1733379 A11 2.91619317 6.29118025 8.71968815
-75.7805818 14.3755458 23.3093829 0.369083041 A12
-0.196645176
-0.732079719
-1.14461356 8.10936356
-1.80803352
-2.69466612 0.0443577624 Notes:
1.
All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the minimum burnup (GWd/MTU) given by the curve fit for the assembly cooling time and initial enrichment. The specific minimum burnup required for each fuel assembly is calculated from the following equation:
Bu = (A1 + A2*En + A3*En2 + A4*En3)* exp [ - (A5 + A6*En + A7*En2 + A8*En3)*Ct ] + A9 + A10*En + A11*En2 + A12*En3 2.
Initial enrichment, En, is the nominal U-235 enrichment. Any enrichment between 1.8 and 4.0 may be used.
3.
Cooling time, Ct, is in years. Any cooling time between 15 years and 25 years may be used. An assembly with a cooling time greater than 25 years must use 25 years.
4.
This Table applies only for pre-extended power uprate (EPU) non-blanketed fuel assemblies. If a non-blanketed assembly is depleted at EPU conditions, none of the burnup accrued at EPU conditions can be credited (i.e., only burnup accrued at pre-EPU conditions may be used as burnup credit).
Delete Table 3.7.14-2 and insert revised Tables 3.7.14-1 through 3.7.14-3
Table 3.7.14-1 (Page 1 of 1)
Pre-EPU Non-Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling time (Ct)
See Notes 1-4 for use of Table 3.7.14-1 Coeff.
Fuel Category I-3 I-4 II-1 II-2 II-3 II-4 II-5 II-6 A1 46.1221
-15.5280
-2.0590 26.4195
-279.3938
-3.7782
-29.0518
-29.0518 A2
-51.4825 13.5960
-2.7964
-23.6884 502.6296 0.7172 38.3795 38.3795 A3 18.4391
-3.4175 3.0982 6.8587
-295.7499 2.1165
-13.3538
-13.3538 A4
-2.0048 0.3637
-0.4715
-0.4980 57.1661
-0.3342 1.6937 1.6937 A5
-0.4998
-1.0368 0.2161
-1.4442
-1.1906 0.1433
-0.4574
-0.4574 A6 0.3474 1.3335
-0.3773 1.6753 1.2161
-0.1589 0.5477 0.5477 A7
-0.0487
-0.4940 0.1893
-0.5777
-0.3067 0.0725
-0.1803
-0.1803 A8 0.0000 0.0574
-0.0265 0.0632 0.0251
-0.0095 0.0184 0.0184 A9
-38.3233
-96.9847 6.7162
-96.0974
-37.9204
-26.9895
-36.7528
-36.7528 A10 24.6155 94.9777
-18.9681 92.2715 31.3948 23.9367 38.4104 38.4104 A11
-3.5675
-28.3931 10.8797
-25.2863
-4.7926
-2.6264
-5.8631
-5.8631 A12 0.3160 3.0898
-1.2782 2.5516 0.3281 0.1421 0.2201 0.2201 Min.
Enrich.
2.00 1.80 1.75 1.55 1.50 1.30 1.15 1.15 Notes:
- 1. All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the minimum burnup (GWd/MTU) given by the curve fit for the assembly cooling time and initial enrichment. The specific minimum burnup required for each fuel assembly is calculated from the following equation. The equation is applicable at enrichments greater than or equal to the value shown as Minimum Enrichment.
Bu = (A1 + A2*En + A3*En2 + A4*En3)
- exp [ - (A5 + A6*En + A7*En2 + A8*En3)*Ct ] + A9 + A10*En + A11*En2 +
A12*En3
- 2. Initial enrichment, En, is the nominal 235U enrichment up to 4.0 wt.%.
- 3. Cooling time, Ct, is in years. Decay (cooling) time credit of 15 years may be used for enrichments less than 2.0 wt.%. Decay (cooling) time credit between 15 and 25 years, inclusive, may be used for any enrichment between 2.0 and 4.0 wt.%, inclusive. An assembly with a cooling time greater than 25 years must use 25 years.
- 4. This table applies only for pre-EPU non-blanketed fuel assemblies. If a non-blanketed assembly is depleted at EPU conditions, none of the burnup accrued at EPU conditions can be credited (i.e., only burnup accrued at pre-EPU conditions may be used as burnup credit).
Spent Fuel Storage 3.7.14
Table 3.7.14-2 (Page 1 of 1)
Mid-Enriched Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling time (Ct)
See Notes 1-4 for use of Table 3.7.14-2 Coeff.
Fuel Category I-3 I-4 II-1 II-2 II-3 II-4 II-5 II-6 A1
-14.0214 0.7356
-10.3764 0.3023
-13.6425
-1.9201
-15.6064 16.2892 A2 11.4137
-1.1927 7.6199
-3.1468 13.5164 2.9502 16.3820
-17.6207 A3
-2.7518 1.4318
-1.2005 2.3278
-2.5923 0.3686
-3.6279 7.2596 A4 0.2743
-0.1832 0.0789
-0.2523 0.1973
-0.0636 0.3114
-0.7399 A5 2.6169
-0.0485 4.8088 0.2364
-0.1211
-0.3267
-0.2816
-0.4164 A6
-2.1487 0.0236
-3.8345
-0.0738 0.1969 0.3766 0.3303 0.5335 A7 0.5878 0.0034 1.0085
-0.0001
-0.0571
-0.1090
-0.0953
-0.1669 A8
-0.0522
-0.0004
-0.0863 0.0016 0.0050 0.0099 0.0087 0.0160 A9
-27.8139
-51.8296
-29.1782
-57.7979
-41.6737
-51.9529
-40.5392
-67.4031 A10 15.7630 41.0704 21.6958 55.4896 42.2351 52.1289 41.5363 74.8527 A11
-0.7370
-8.3986
-3.2089
-13.5089
-8.9287
-11.9184
-8.8545
-19.0424 A12
-0.0324 0.7265 0.2488 1.2360 0.7680 1.0595 0.7866 1.7507 Min.
Enrich.
2.00 1.75 1.75 1.55 1.35 1.30 1.30 1.15 Notes:
- 1. All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the minimum burnup (GWd/MTU) given by the curve fit for the assembly cooling time and initial enrichment. The specific minimum burnup required for each fuel assembly is calculated from the following equation. The equation is applicable at enrichments greater than or equal to the value shown as Minimum Enrichment.
Bu = (A1 + A2*En + A3*En2 + A4*En3)
- exp [ - (A5 + A6*En + A7*En2 + A8*En3)*Ct ] + A9 + A10*En + A11*En2 +
A12*En3
- 2. Initial enrichment, En, is the nominal 235U enrichment up to 5.0 wt.%. Axial blanket material is not considered when determining enrichment.
- 3. Cooling time, Ct, is in years. No decay (cooling) time credit may be used for enrichments less than 2.0 wt.%.
Decay (cooling) time credit between 0 and 25 years, inclusive, may be used for any enrichment between 2.0 and 5.0 wt.%, inclusive. An assembly with a cooling time greater than 25 years must use 25 years.
- 4. This table applies only for assemblies with a blanket enULFKPHQW wt% 235U.
Spent Fuel Storage 3.7.14
Table 3.7.14-3 (Page 1 of 1)
Post-EPU Non-Blanketed Fuel - Coefficients to Calculate the Minimum Required Fuel Assembly Burnup (Bu) as a Function of Enrichment (En) and Cooling time (Ct)
See Notes 1-4 for use of Table 3.7.14-3 Coeff.
Fuel Category I-3 I-4 II-1 II-2 II-3 II-4 II-5 II-6 A1 32.2479
-4.1991 12.9596
-8.7984 1.2361
-13.6999
-4.4636 16.9460 A2
-32.5873 0.5751
-16.0005 17.5883 3.9352 13.5880 4.4226
-16.9514 A3 10.8045 1.2741 6.3237
-6.8331
-1.1864
-2.6470 0.3955 6.7299 A4
-1.0774
-0.1682
-0.6838 0.8117 0.1753 0.2090
-0.0894
-0.6660 A5
-0.9953
-0.9249
-0.5872 0.0832 0.0667 0.2213
-0.2197
-0.4412 A6 0.9362 0.8428 0.5836
-0.1491
-0.1430
-0.1129 0.2649 0.5695 A7
-0.2713
-0.2310
-0.1721 0.0770 0.0840 0.0290
-0.0800
-0.1805 A8 0.0260 0.0205 0.0169
-0.0095
-0.0112
-0.0025 0.0078 0.0175 A9
-55.7079
-31.2188
-30.7329
-33.6356
-42.2030
-34.7146
-53.7542
-64.6698 A10 40.9920 22.8793 22.0019 18.4614 34.1725 34.4020 56.0845 70.4969 A11
-8.4183
-2.8703
-3.2299 0.7440
-4.6731
-6.5830
-13.3110
-17.5951 A12 0.7732 0.1971 0.2932
-0.2665 0.2560 0.6009 1.2772 1.6628 Min.
Enrich.
2.00 1.75 1.75 1.55 1.35 1.30 1.30 1.15 Notes:
- 1. All relevant uncertainties are explicitly included in the criticality analysis. For instance, no additional allowance for burnup uncertainty or enrichment uncertainty is required. For a fuel assembly to meet the requirements of a Fuel Category, the assembly burnup must exceed the minimum burnup (GWd/MTU) given by the curve fit for the assembly cooling time and initial enrichment. The specific minimum burnup required for each fuel assembly is calculated from the following equation. The equation is applicable at enrichments greater than or equal to the value shown as Minimum Enrichment.
Bu = (A1 + A2*En + A3*En2 + A4*En3)
- exp [ - (A5 + A6*En + A7*En2 + A8*En3)*Ct ] + A9 + A10*En + A11*En2 +
A12*En3
- 2. Initial enrichment, En, is the nominal 235U enrichment up to 5.0 wt.%.
- 3. Cooling time, Ct, is in years. No decay (cooling) time credit may be used for enrichments less than 2.0 wt.%.
Decay (cooling) time credit between 0 and 25 years, inclusive, may be used for any enrichment between 2.0 and 5.0 wt.%, inclusive. An assembly with a cooling time greater than 25 years must use 25 years.
- 4. This table applies for all post-EPU non-blanketed assemblies.
Spent Fuel Storage 3.7.14
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-5 Amendment Nos. 297 and 290 Table 3.7.14-3 (page 1 of 1)
Fuel Categories Ranked by Reactivity See Notes 1-5 for use of Table 3.7.14-3 Region I I-1 High Reactivity Low Reactivity I-2 I-3 I-4 Region II II-1 High Reactivity Low Reactivity II-2 II-3 II-4 II-5 Notes:
- 1. Fuel Category is ranked by decreasing order of reactivity without regard for any reactivity-reducing mechanisms, e.g., Category I-2 is less reactive than Category I-1, etc. The more reactive fuel categories require compensatory measures to be placed in Regions I and II of the spent fuel pit, e.g., use of water filled cells, Metamic inserts, or full length RCCAs.
- 2. Any higher numbered fuel category can be used in place of a lower numbered fuel category from the same Region.
- 3. Category I-1 is fresh unburned fuel up to 5.0 wt% U-235 enrichment.
- 4. Category I-2 is fresh unburned fuel that obeys the Integral Fuel Burnable Adsorber (IFBA) requirements of Table 3.7.14-4.
- 5. All Categories except I-1 and I-2 are determined from Tables 3.7.14-1 and 3.7.14-2.
3.7.14-4 Add a row at the bottom of Region II, place "II-6" in middle column, and shift "Low Reactivity" down to align with this new II-6 row.
LCO 3.7.14.b.6.
3.7.14-1 through 3.7.14-3
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-6 Amendment Nos. 297 and 290 Table 3.7.14-4 (page 1 of 1)
IFBA Requirements for Fuel Category I-2 Nominal Enrichment (wt% U-235)
Minimum Required Number of IFBA Pins (QU
0
(QU
32
(QU
64
(QU
80 Delete Table 3.7.14-4
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-7 Amendment Nos. 297 and 290 Figure 3.7.14-1 (page 1 of 1)
Allowable Region I Storage Arrays See Notes 1-8 for use of Figure 3.7.14-1 DEFINITION ILLUSTRATION Array I-A Checkerboard pattern of Category I-1 assemblies and empty (water-filled) cells.
Array I-B Category I-4 assembly in every cell.
Array I-C Combination of Category I-2 and I-4 assemblies. Each Category I-2 assembly shall contain a full length RCCA.
Array I-D Category I-3 assembly in every cell. One of every four assemblies contains a full length RCCA.
Notes:
- 1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.
- 2. Category I-1 is fresh unburned fuel up to 5.0 wt% U-235 enrichment.
- 3. Category I-2 is fresh unburned fuel that obeys the IFBA requirements in Table 3.7.14-4.
- 4. Categories I-3 and I-4 are determined from Tables 3.7.14-1 and 3.7.14-2.
- 5. Shaded cells indicate that the fuel assembly contains a full length RCCA.
- 6. X indicates an empty (water-filled) cell.
- 7. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.
- 8. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.
I-1 X
X I-1 I-4 I-4 I-4 I-4 I-2 I-4 I-2 I-2 I-4 I-4 I-4 I-4 I-2 I-2 I-2 I-2 I-2 I-4 I-2 I-2 I-3 I-3 I-3 I-3 LCO 3.7.14.b.6.
3.7.14-1 through 3.7.14-3
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-8 Amendment Nos. 297 and 290 Figure 3.7.14-2 (page 1 of 1)
Allowable Region II Storage Arrays See Notes 1-6 for use of Figure 3.7.14-2 DEFINITION ILLUSTRATION Array II-A Category II-1 assembly in three of every four cells; one of every four cells is empty (water-filled);
the cell diagonal from the empty cell contains a Metamic insert or full length RCCA.
Array II-B Checkerboard pattern of Category II-3 and II-5 assemblies With two of every four cells containing a Metamic insert or full length RCCA.
Array II-C Category II-4 assembly in every cell with two of every four Cells containing a Metamic insert or full length RCCA.
Array II-D Category II-2 assembly in every cell with three of every four Cells containing a Metamic insert or full length RCCA.
Notes:
- 1. In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.
- 2. Fuel categories are determined from Tables 3.7.14-1 and 3.7.14-2.
- 3. Shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.
- 4. X indicates an empty (water-filled) cell.
- 5. Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.
- 6. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.
II-1 II-1 X
II-1 X
II-1 II-1 II-1 II-3 II-5 II-3 II-5 II-5 II-3 II-5 II-3 II-4 II-4 II-4 II-4 II-4 II-4 II-4 II-4 II-2 II-2 II-2 II-2 Un-capitalize "with" and "cells" Array II-E Category II-6 assembly in every cell with one of every four cells containing a Metamic insert or full length RCCA.
3.7.14-1 through 3.7.14-3
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-9 Amendment Nos. 297 and 290 Figure 3.7.14-3 (page 1 of 1)
Interface Restrictions Between Region I and Region II Arrays See Notes 1-8 for use of Figure 3.7.14-3 DEFINITION ILLUSTRATION Array II-A, as defined in Figure 3.7.14-2, when placed on The interface with Region I shall have the empty cell in the row adjacent to the Region I rack.
Arrays II-B, II-C and II-D, as defined in Figure 3.7.14-2, when placed on the interface with Region I shall have an insert in every cell in the row adjacent to the Region I rack.
Notes:
1.
In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.
2.
Fuel categories are determined from Tables 3.7.14-1 and 3.7.14-2.
3.
Shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.
4.
X indicates an empty (water-filled) cell.
5.
Attributes for each 2x2 array are as stated in the definition. Diagram is for illustrative purposes only.
Region I Array I-B is depicted as the example; however, any Region I array is allowed provided that
- a. For Array I-D, the RCCA shall be in the row adjacent to the Region II rack, and
- b. Array I-A shall not interface with Array II-D.
6.
If no fuel is stored adjacent to Region II in Region I, then the interface restrictions are not applicable.
7.
Figure 3.7.14-3 is applicable only to the Region I - Region II interface. There are no restrictions for the interfaces with the cask area rack.
8.
An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.
Region I Rack I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 II-1 X
II-1 X
II-1 II-1 II-1 II-1 Array II-A Region I Rack Region I Rack Region I Rack I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 II-3 II-5 II-3 II-5 II-4 II-4 II-4 II-4 II-2 II-2 II-2 II-2 II-5 II-3 II-5 II-3 II-4 II-4 II-4 II-4 II-2 II-2 II-2 II-2 Array II-B Array II-C Array II-D Delete Figure 3.7.14-3 and insert revised Figure 3.7.14-3
Figure 3.7.14-3 (page 1 of 2)
Interface Restrictions Between Region I and Region II Arrays See Notes 1-13 for use of Figure 3.7.14-3 Array I-A - Region II Array I-A Array I-A Array I-A X
I-1 X
I-1 X
I-1 X
I-1 X
I-1 X
I-1 I-1 X
I-1 X
I-1 X
I-1 X
I-1 X
I-1 X
II-1 X
II-1 X
II-4 II-4 II-4 II-4 II-6 II-6 II-6 II-6 II-1 II-1 II-1 II-1 II-4 II-4 II-4 II-4 II-6 II-6 II-6 II-6 Array II-A Array II-C Array II-E Array I-B - Region II Array I-B Array I-B I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 I-4 II-1 X
II-1 X
II-2 II-2 II-2 II-2 II-1 II-1 II-1 II-1 II-2 II-2 II-2 II-2 Array II-A Array II-D Array I-D - Region II Array I-D Array I-D Array I-D Array I-D Array I-D I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 II-1 X
II-1 X
II-3 II-5 II-3 II-5 II-3 II-5 II-3 II-5 II-3 II-5 II-3 II-5 II-4 II-4 II-4 II-4 II-1 II-1 II-1 II-1 II-5 II-3 II-5 II-3 II-5 II-3 II-5 II-3 II-5 II-3 II-5 II-3 II-4 II-4 II-4 II-4 Array II-A Array II-B Array II-B Array II-B Array II-C Array I-D Array I-D Array I-D Array I-D I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 I-3 II-4 II-4 II-4 II-4 II-4 II-4 II-4 II-4 II-2 II-2 II-2 II-2 II-2 II-2 II-2 II-2 II-4 II-4 II-4 II-4 II-4 II-4 II-4 II-4 II-2 II-2 II-2 II-2 II-2 II-2 II-2 II-2 Array II-C Array II-C Array II-D Array II-D Spent Fuel Storage 3.7.14
Figure 3.7.14-3 (page 2 of 2)
Interface Restrictions Between Region I and Region II Arrays See Notes 1-13 for use of Figure 3.7.14-3 Notes:
In all arrays, an assembly of lower reactivity can replace an assembly of higher reactivity.
Fuel categories are determined from Tables 3.7.14-1 through 3.7.14-3.
Region, shaded cells indicate that the fuel assembly contains a full length RCCA.
Region,, shaded cells indicate that the cell contains a Metamic insert or the fuel assembly
contains a full length RCCA.
X indicates an empty (water-filled) cell.
Region I and Region II storage cells do not necessarily align across the interface as shownin the figure. There are no restrictions associated with cell alignment across the interface.
If no fuel is stored adjacent to Region II in Region I, then the interface restrictions are not
applicable.
Array I-A is subject to the following restrictions:
D
Array I-A shall not interface with Array II-B or II-D.
E
Array I-A - Array II-A Interface: The Array II-A empty cell shall be placed on the
interface.
F
Array I-A - Array II-C Interface: All required Metamic inserts or RCCAs must be
placed along the interface.
G
Array I-A - Array II-E Interface: All required Metamic inserts or RCCAs must be
placed along the interface.
Array I-B is subject to the following restrictions:
D
Array I-B - Array II-A Interface: The Array II-A empty cell shall be placed on the
interface.
E
Array I-B - Array II-D Interface: The Array II-D assemblies on the interface must all
contain Metamic inserts or RCCAs.
F
There are no restrictions for Arrays II-B, II-C or II-E with Array I-B.
The same restrictions noted for Array I-B apply to Array I-C, with no additional restrictions on
the Region I side regarding Fuel Categories I-2 and I-4.
Array I-D is subject to the following restrictions:
D
Array I-D - Array II-A Interface: The Array II-A empty cell shall be placed on the
interface.
E
Array I-D - Array II-B Interface: A storage cell on the interface in each storage array
must contain a Metamic insert or RCCA on at least one side of the interface.
F
Array I-D - Array II-C Interface: A storage cell on the interface in each storage array
must contain a Metamic insert or RCCA on at least one side of the interface.
G
Array I-D - Array II-D Interface: Either all Array II-D cells on the interface must
contain Metamic inserts or RCCAs, or each Region I storage array and each
RegionII storage array must contain a Metamic insert or RCCA in a storage cell on
theinterface.
H
Array I-D - Array II-E Interface: No interface restrictions.
This Figure is only applicable to the Region I - Region II interface. There are no restrictions
for the interfaces with the Cask Area Rack.
An empty (water-filled) cell may be substituted for any fuel containing cell in all storage
arrays.
Spent Fuel Storage 3.7.14
Design Features 4.0 Turkey Point Unit 3 and Unit 4 4.0-1 Amendment Nos. 297 and 290 4.0 DESIGN FEATURES 4.1 Site Location The site is approximately 25 miles south of Miami, 8 miles east of Florida City and 9 miles southeast of Homestead, Florida.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy-4, ZIRLO, or Optimized ZIRLO' fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material.
Limited substitutions of stainless steel filler rods for fuel rods, or by vacant rod positions, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods.
4.2.2 Control Rod Assemblies The reactor core shall contain 45 control rod assemblies. The control material shall be silver indium cadmium, as approved by the NRC.
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent, b.
keff 0.95 if fully flooded with water borated to 500 ppm, which includes an allowance for biases and uncertainties as described in Section 9.5 of the UFSAR, c.
keff 1.0 if fully flooded with unborated water, which includes an allowance for biases and uncertainties as described in Section 9.5 of the UFSAR, d.
A nominal 10.6 inch center to center distance between fuel assemblies placed in Region I of the fuel storage racks, e.
A nominal 9.0 inch center to center distance between fuel assemblies placed in Region II of the fuel storage racks, 550
Design Features 4.0 Turkey Point Unit 3 and Unit 4 4.0-2 Amendment Nos. 297 and 290 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) f.
A nominal 10.1 inch center to center distance in the east-west direction and a nominal 10.7 inch center to center distance in the north-south direction for the cask area storage rack, 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum U-235 enrichment of 4.5 weight percent if the assemblies contain no burnable adsorber
- rods, b.
Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent if the assemblies contain at least 16 integral fuel burnable adsorber rods.
c.
A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks to assure keff 0.98 for optimum moderation conditions, and d.
A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks to assure keff 0.95 for fully flooded conditions.
4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below a level of 6 feet above the fuel assemblies in the storage racks.
4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1535 fuel assemblies.
This page is for information only. No changes are proposed for this page.
Turkey Point Nuclear Plant L-2023-077 Docket Nos. 50-250 and 50-251 ATTACHMENT 2 Turkey Point Technical Specifications Bases Page Markups (6 pages follow)
Fuel Storage Pool Boron Concentration B 3.7.13 Turkey Point Unit 3 and Unit 4 B 3.7.13-1 Revision No. 0 B 3.7 PLANT SYSTEMS B 3.7.13 Fuel Storage Pool Boron Concentration BASES BACKGROUND The spent fuel storage racks provide safe subcritical storage of fuel assemblies by providing sufficient center-to-center spacing or a combination of spacing and poison to assure: a) keff less than or equal to 0.95 with a minimum soluble boron concentration of 500 ppm present, and b) keff less than 1.0 when flooded with unborated water for normal operations and postulated accidents. The 500 ppm value is needed to assure keff less than 0.95 for normal operating conditions. The criticality analysis needs 1700 ppm to assure keff less than 0.95 under the worst case accident condition. There is significant margin between the calculated ppm requirement and the spent fuel boron concentration requirement of 2300 ppm. The higher boron concentration value is chosen because, during refueling, the water volume in the spent fuel pool, the transfer canal, the refueling canal, the refueling cavity, and the reactor vessel form a single mass.
The spent fuel racks are divided into two regions, Region I and Region II.
The Region I permanent racks have a 10.6 inch center-to-center spacing.
The Region lI racks have a 9.0 inch center-to-center spacing. The cask area storage rack has a nominal 10.1 inch center-to-center spacing in the east-west direction and a nominal 10.7 inch center-to-center spacing in the north-south direction.
Any fuel for use at Turkey Point, and enriched to less than or equal to 5.0 wt% U-235, may be stored in the cask area storage rack. Fresh or irradiated fuel assemblies not stored in the cask area storage rack shall be stored in accordance with LCO 3.7.14.
Fresh unirradiated fuel may be placed in the permanent Region I racks in accordance with the restrictions of Figure 3.7.14-1. Prior to placement of irradiated fuel in Region I or II spent fuel storage rack cell locations, strict controls are employed to evaluate burnup of the fuel assembly. Upon determination that the fuel assembly meets the nominal burnup and associated post-irradiation cooling time requirements of Table 3.7.14-1 or Table 3.7.14-2, it may be placed in a Region I or II cell in accordance with the restrictions of Figures 3.7.14-1 through 3.7.14-3, respectively.
For all assemblies with blanketed fuel, the initial enrichment is based on the central zone enrichment (i.e., between the axial blankets) consistent with the assumptions of the analysis. These positive controls assure that the fuel enrichment limits, burnup, and post irradiation cooling time requirements assumed in the safety analyses will not be violated.
550 2350 The boron concentration value remains bounding during refueling while Tables 3.7.14-1 through 3.7.14-3,
Fuel Storage Pool Boron Concentration B 3.7.13 Turkey Point Unit 3 and Unit 4 B 3.7.13-2 Revision No. 0 BASES BACKGROUND (continued)
The water in the spent fuel storage normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting keff of 0.95 be evaluated in the absence of soluble boron. Hence, the design of both regions is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the regions fully loaded. The double contingency principle discussed in ANSI/ANS-8.1-1983 (Ref. 1) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. Prior to movement of an assembly, it is necessary to perform SR 3.7.13.1.
APPLICABLE The following fuel handling accidents are evaluated to ensure that no SAFETY hazards are created: a) A fuel assembly is dropped in containment. b) A ANALYSES spent fuel cask is dropped in the passage between the spent fuel pits of Units 3 & 4 while transferring a fuel element between the spent fuel pits.
The consideration of a cask drop accident is historical and is retained as discussed in UFSAR Section 14.2.1.3. (Ref.2)
Since the spent fuel cask will not be handled over or in the vicinity of spent fuel except as provided for in UFSAR Section 14.2.1.3.1, the re-racking does not result in a significant increase in the probability of the cask drop accident previously evaluated in the Turkey Point UFSAR.
Furthermore, as shown in UFSAR Section 14.2.1.3.2, by requiring the decay time of spent fuel to be a minimum of 1525 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.802625e-4 months <br /> prior to moving a spent fuel cask into the spent fuel pit, the potential offsite doses will be well within 10 CFR Part 100 limits should a dropped cask strike the stored fuel assemblies. The proposed spent fuel pit modifications will not increase the radiological consequences of a cask drop accident previously evaluated (Ref.2).
The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The fuel storage pool boron concentration is required to be 2300 ppm.
The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference 2. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel storage pool.
2350 1.0
Fuel Storage Pool Boron Concentration B 3.7.13 Turkey Point Unit 3 and Unit 4 B 3.7.13-3 Revision No. 0 BASES APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel storage, until a complete spent fuel storage verification has been performed following the last movement of fuel assemblies in the spent fuel storage. This LCO does not apply following the verification, since the verification would confirm that there are no misloaded fuel assemblies.
With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.
ACTIONS A.1, A.2.1, and A.2.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.
When the concentration of boron in fuel storage is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. Alternatively, beginning a verification of the fuel storage pool fuel locations, to ensure proper locations of the fuel, can be performed. However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.
If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.13.1 REQUIREMENTS This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES 1.
Double contingency principle of ANSI/ANS-8.1-1983.
2Property "ANSI code" (as page type) with input value "ANSI/ANS-8.1-1983.</br></br>2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
UFSAR, Section 14.2.1.
This page is for information only. No changes are proposed for this page.
Spent Fuel Storage B 3.7.14 Turkey Point Unit 3 and Unit 4 B 3.7.14-1 Revision No. 0 B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Storage BASES BACKGROUND The spent fuel pit storage racks provide safe subcritical storage of fuel assemblies by providing sufficient center-to-center spacing or a combination of spacing and poison to assure: a) keff less than or equal to 0.95 with a minimum soluble boron concentration of 500 ppm present, and b) keff less than 1.0 when flooded with unborated water for normal operations and postulated accidents. The 500 ppm value is needed to assure keff less than 0.95 for normal operating conditions. The criticality analysis needs 1700 ppm to assure keff less than 0.95 under the worst case accident condition. There is significant margin between the calculated ppm requirement and the spent fuel boron concentration requirement of 2300 ppm. The higher boron concentration value is chosen because, during refueling, the water volume in the spent fuel pit, the transfer canal, the refueling canal, the refueling cavity, and the reactor vessel form a single mass.
The spent fuel pit storage racks are divided into two regions, Region I and Region II. The Region I permanent racks have a 10.6 inch center-to-center spacing. The Region lI racks have a 9.0 inch center-to-center spacing. The cask area storage rack has a nominal 10.1 inch center-to-center spacing in the east-west direction and a nominal 10.7 inch center-to-center spacing in the north-south direction.
Any fuel for use at Turkey Point, and enriched to less than or equal to 5.0 wt% U-235, may be stored in the cask area storage rack. Fresh or irradiated fuel assemblies not stored in the cask area storage rack shall be stored in accordance with LCO 3.7.14.
Fresh unirradiated fuel may be placed in the permanent Region I racks in accordance with the restrictions of Figure 3.7.14-1. Fresh unirradiated fuel may be placed in the permanent Region II racks in accordance with the restrictions of Figure 3.7.14-2. Prior to placement of irradiated fuel in Region I or II spent fuel pit storage rack cell locations, strict controls are employed to evaluate burnup of the fuel assembly. Upon determination that the fuel assembly meets the nominal burnup and associated post-irradiation cooling time requirements of Table 3.7.14-1 or Table 3.7.14-2, it may be placed in a Region I or II cell in accordance with the restrictions of Figures 3.7.14-1 through 3.7.14-3, respectively.
For all assemblies with blanketed fuel, the initial enrichment is based on the central zone enrichment (i.e., between the axial blankets) consistent with the assumptions of the analysis. These positive controls assure that the fuel enrichment limits, burnup, and post irradiation cooling time requirements assumed in the safety analyses will not be violated.
550 2350 The boron concentration value remains bounding during refueling while Tables 3.7.14-1 through 3.7.14-3,
Spent Fuel Storage B 3.7.14 Turkey Point Unit 3 and Unit 4 B 3.7.14-2 Revision No. 0 BASES BACKGROUND (continued)
The water in the spent fuel pit normally contains soluble boron, which results in large subcriticality margins under actual operating conditions.
However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting keff of 1.0 be evaluated in the absence of soluble boron. Hence, the design of both regions is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the regions fully loaded. The double contingency principle discussed in ANSI/ANS-18.1-1983 (Ref. 1) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. Prior to movement of an assembly, it is necessary to perform SR 3.7.13.1.
APPLICABLE The hypothetical accidents can only take place during or as a result of SAFETY the movement of an assembly (Ref. 2). For these accident occurrences, ANALYSES the presence of soluble boron in the spent fuel pit (controlled by LCO 3.7.13, "Fuel Storage Pool Boron Concentration") prevents criticality in both regions. By closely controlling the movement of each assembly and by checking the location of each assembly after movement, the time period for potential accidents may be limited to a small fraction of the total operating time. During the remaining time period with no potential for accidents, the operation may be under the auspices of the accompanying LCO.
The configuration of fuel assemblies in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO The restrictions on the placement of fuel assemblies within the spent fuel pit, in accordance with Figure 3.7.14-1 through 3.7.14-3 and Tables 3.7.14-1 through 3.7.14-4, in the accompanying LCO, ensures the keff of the spent fuel pit will always remain < 1.0, assuming the pool to be flooded with unborated water.
The LCO states that there are no restrictions on storage of fresh or irradiated fuel assemblies in the cask area storage rack. This is because in the cask area rack criticality is prevented by the design of the rack which limits fuel assembly interaction by fixing the separation distance between stored assemblies and/or by placing a neutron absorber panel between storage cells.
APPLICABILITY This LCO applies whenever any fuel assembly is stored in Region I or Region II of the spent fuel pit.
are discussed in Ref. 2.
Spent Fuel Storage B 3.7.14 Turkey Point Unit 3 and Unit 4 B 3.7.14-3 Revision No. 0 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.
When the configuration of fuel assemblies stored in the spent fuel pit is not in accordance with Figure 3.7.14-1, Figure 3.7.14-2, or Figure 3.7.14-3, the immediate action is to initiate action to make the necessary fuel assembly movement(s) to bring the configuration into compliance.
If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies by administrative means that fuel assemblies stored in Regions I and II are stored in accordance with the requirements of Figure 3.7.14-1 through Figure 3.7.14-3 with credit for burnup and cooling time taken in determining acceptable placement locations for spent fuel in the two region spent fuel racks. For fuel assemblies not meeting requirements of Figures 3.7.14-1 through 3.7.14-3, performance of this SR will ensure compliance with Specification 4.3.1.1.
REFERENCES 1.
Double contingency principle of ANSI/ANS-8.1-1983.
2Property "ANSI code" (as page type) with input value "ANSI/ANS-8.1-1983.</br></br>2" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
UFSAR, Section 14.2.1.
from Tables 3.7.14-1 through 3.7.14-3