ML24319A294
| ML24319A294 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 12/11/2024 |
| From: | Perry Buckberg Plant Licensing Branch II |
| To: | Coffey B Florida Power & Light Co |
| Buckberg, P. NRR/DORL 415-1383 | |
| References | |
| EPID L-2022-LLA-0128 | |
| Download: ML24319A294 (1) | |
Text
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION December 11, 2024 Robert Coffey Executive Vice President, Nuclear and Chief Nuclear Officer Florida Power & Light Company Mail Stop: EX/JB 700 Universe Blvd.
Juno Beach, FL 33408
SUBJECT:
TURKEY POINT NUCLEAR GENERATING, UNIT NOS. 3 AND 4 - ISSUANCE OF AMENDMENT NOS. 299 AND 292 REGARDING REVISING THE SPENT FUEL POOL CRITICALITY ANALYSIS CONSISTENT WITH THE TRANSITION TO A 24-MONTH FUEL CYCLE (EPID L-2023-LLA-0142)
Dear Robert Coffey:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 299 to Subsequent Renewed Facility Operating License No. DPR-31 and Amendment No. 292 to Subsequent Renewed Facility Operating License No. DPR-41 for Turkey Point Nuclear Generating, Unit Nos. 3 and 4, respectively. These amendments revise technical specifications (TSs) to allow for an updated spent fuel pool criticality safety analysis in response to your application dated October 11, 2023, as supplemented by letters dated June 21, 2024, September 18, 2024, September 24, 2024, October 15, 2024, October 22, 2024, October 29, 2024, and November 12, 2024.
Specifically, the amendments revise the TSs 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 that accounts for the impact on the spent fuel from a proposed transition to 24-month fuel cycles.
to this letter contains Proprietary Information. When separated from Enclosure 3, this letter is DECONTROLLED.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION A copy of the related safety evaluation is also enclosed. Notice of issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Perry H. Buckberg, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-250 and 50-251
Enclosures:
- 1. Amendment No. 299 to DPR-31
- 2. Amendment No. 292 to DPR-41
- 3. Safety Evaluation (Proprietary)
- 4. Safety Evaluation (Non-Proprietary) cc: Listserv
FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-250 TURKEY POINT NUCLEAR GENERATING, UNIT NO. 3 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 299 Subsequent Renewed License No. DPR-31
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Florida Power & Light Company (the licensee) dated October 11, 2023, as supplemented by letters dated June 21, 2024, September 18, 2024, September 24, 2024, October 15, 2024, October 22, 2024, October 29, 2024, and November 12, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-31 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 299, are hereby incorporated into this subsequent renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented no later than the Unit No. 3 spring 2026 reload campaign.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Subsequent Renewed Facility Operating License and Technical Specifications Date of Issuance: December 11, 2024
FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-251 TURKEY POINT NUCLEAR GENERATING, UNIT NO. 4 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 292 Subsequent Renewed License No. DPR-41
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Florida Power & Light Company (the licensee) dated October 11, 2023, as supplemented by letters dated June 21, 2024, September 18, 2024, September 24, 2024, October 15, 2024, October 22, 2024, October 29, 2024, and November 12, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-41 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 292 are hereby incorporated into this subsequent renewed operating license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this subsequent renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented no later than the Unit No. 4 spring 2025 reload campaign.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Subsequent Renewed Facility Operating License and Technical Specifications Date of Issuance: December 11, 2024
ATTACHMENT TO LICENSE AMENDMENT NOS. 299 AND 292 TURKEY POINT NUCLEAR GENERATING, UNIT NOS. 3 AND 4 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NOS. DPR-31 AND DPR-41 DOCKET NOS. 50-250 AND 50-251 Replace the following pages of the Subsequent Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert DPR-31, page 3 DPR-31, page 3 DPR-41, page 3 DPR-41, page 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert 3.7.13-1 3.7.13-1 3.7.14-1 3.7.14-1 3.7.14-2 3.7.14-2 3.7.14-3 3.7.14-3 3.7.14-4 3.7.14-4 3.7.14-5 3.7.14-5 3.7.14-6 3.7.14-6 3.7.14-7 3.7.14-7 3.7.14-8 3.7.14-8 3.7.14-9 3.7.14-9 3.7.14-10 4.0-1 4.0-1 4.0-2 4.0-2
3 Subsequent Renewed License No. DPR-31 Amendment No. 29
applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified below:
$
Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2644 megawatts (thermal).
%
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. , are hereby incorporated into this subsequent renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
&
Deleted.
'
Fire Protection FPL shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment requests dated June 28, 2012 and October 17, 2018 (and supplements dated September 19, 2012; March 18, April 16, and May 15, 2013; January 7, April 4, June 6, July 18, September 12, November 5, and December 2, 2014; and February 18, 2015; October 24, and December 3, 2018; and January 31, 2019), and as approved in the safety evaluations dated May 28, 2015 and March 27, 2019. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the
3 Subsequent Renewed License No. DPR-41 Amendment No. 29
$
Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2644 megawatts (thermal).
%
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. are hereby incorporated into this subsequent renewed operating license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this subsequent renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
&
Deleted
'
Fire Protection FPL shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment requests dated June 28, 2012 and October 17, 2018 (and supplements dated September 19, 2012; March 18, April 16, and May 15, 2013; January 7, April 4, June 6, July 18, September 12, November 5, and December 2, 2014; and February 18, 2015; October 24, and December 3, 2018; and January 31, 2019), and as approved in the safety evaluations dated May 28, 2015 and March 27, 2019. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the
Fuel Storage Pool Boron Concentration 3.7.13 Turkey Point Unit 3 and Unit 4 3.7.13-1 Amendment Nos. 29 and 29
3.7 PLANT SYSTEMS 3.7.13 Fuel Storage Pool Boron Concentration LCO 3.7.13 The fuel storage pool boron concentration shall be 2350 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
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-1 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 through 3.7.14-3, as applicable. The reactivity rank of fuel assemblies in the 2x2 array (rank determined using Table 3.7.14-4) 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 through 3.7.14-3, as applicable. The reactivity rank of fuel assemblies in the 2x2 array (rank determined using Table 3.7.14-4) 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.
Amendment Nos. 29 and 29
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-2 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.
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.
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 Amendment Nos. 29 and 29
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 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 26.4195
-3.7782
-29.0518
-29.0518 A2
-51.4825 13.5960
-2.7964
-23.6884
-23.6884 0.7172 38.3795 38.3795 A3 18.4391
-3.4175 3.0982 6.8587 6.8587 2.1165
-13.3538
-13.3538 A4
-2.0048 0.3637
-0.4715
-0.4980
-0.4980
-0.3342 1.6937 1.6937 A5
-0.4998
-1.0368 0.2161
-1.4442
-1.4442 0.1433
-0.4574
-0.4574 A6 0.3474 1.3335
-0.3773 1.6753 1.6753
-0.1589 0.5477 0.5477 A7
-0.0487
-0.4940 0.1893
-0.5777
-0.5777 0.0725
-0.1803
-0.1803 A8 0.0000 0.0574
-0.0265 0.0632 0.0632
-0.0095 0.0184 0.0184 A9
-38.3233
-96.9847 6.7162
-96.0974
-96.0074
-26.9895
-36.7528
-36.7528 A10 24.6155 94.9777
-18.9681 92.2715 92.2715 23.9367 38.4104 38.4104 A11
-3.5675
-28.3931 10.8797
-25.2863
-25.2863
-2.6264
-5.8631
-5.8631 A12 0.3160 3.0898
-1.2782 2.5516 2.5516 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).
Amendment Nos. 29 and 29
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-4 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
-51.9429
-40.4692
-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 enrichment 2.6 wt% 235U.
Amendment Nos. 29 and 29
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-5 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.
Amendment Nos. 29 and 29
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-6 Table 3.7.14-4 (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 II-6 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 LCO 3.7.14.b.6.
5.
All Categories except I-1 and I-2 are determined from Tables 3.7.14-1 through 3.7.14-3.
Amendment Nos. 29 and 29
Spent Fuel Storage 3.7.14 Turkey Point Unit 3 and Unit 4 3.7.14-10 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:
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 through 3.7.14-3.
3.
Region I shaded cells indicate that the fuel assembly contains a full length RCCA.
4.
Region II shaded cells indicate that the cell contains a Metamic insert or the fuel assembly contains a full length RCCA.
5.
X indicates an empty (water-filled) cell.
6.
Region I and Region II storage cells do not necessarily align across the interface as shown in the figure. There are no restrictions associated with cell alignment across the interface.
7.
If no fuel is stored adjacent to Region II in Region I, then the interface restrictions are not applicable.
8.
Array I-A is subject to the following restrictions:
a.
Array I-A shall not interface with Array II-B or II-D.
b.
Array I-A - Array II-A Interface: The Array II-A empty cell shall be placed on the interface.
c.
Array I-A - Array II-C Interface: All required Metamic inserts or RCCAs must be placed along the interface.
d.
Array I-A - Array II-E Interface: All required Metamic inserts or RCCAs must be placed along the interface.
9.
Array I-B is subject to the following restrictions:
a.
Array I-B - Array II-A Interface: The Array II-A empty cell shall be placed on the interface.
b.
Array I-B - Array II-D Interface: The Array II-D assemblies on the interface must all contain Metamic inserts or RCCAs.
c.
There are no restrictions for Arrays II-B, II-C or II-E with Array I-B.
- 10. 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.
- 11. Array I-D is subject to the following restrictions:
a.
Array I-D - Array II-A Interface: The Array II-A empty cell shall be placed on the interface.
b.
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.
c.
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.
d.
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 Region II storage array must contain a Metamic insert or RCCA in a storage cell on the interface.
e.
Array I-D - Array II-E Interface: No interface restrictions.
- 12. This Figure is only applicable to the Region I - Region II interface. There are no restrictions for the interfaces with the Cask Area Rack.
- 13. An empty (water-filled) cell may be substituted for any fuel containing cell in all storage arrays.
Amendment Nos. 29 and 29
Design Features 4.0 Turkey Point Unit 3 and Unit 4 4.0-1 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 550 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, Amendment Nos. 29 and 29
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Amendment Nos. 29 and 29
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION selected fuel assemblies, as well as for the presence of Integrated Fuel Burnable Absorber (IFBA) rods in fresh fuel evaluations. Credit is also taken for the negative reactivity associated with burnup and post-irradiation cooling time. FPL submitted Westinghouse Report, WCAP-18830-P, Revision 0, Turkey Point Fuel Storage Criticality Analysis for 24 Month Cycles, dated September 22, 2023 (ML23265A548 - Package), documenting the SFP criticality safety analysis for Turkey Point.
The supplements dated June 21, 2024, September 18, 2024, October 15, 2024, and October 29, 2024, provided additional information that corrected inconsistencies and calculational results. The supplements dated September 18, 2024, October 22, 2024, and November 12, 2024, provided additional information in response to NRC staff requests for additional information (RAIs). These supplements did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on January 23, 2024 (89 FR 4342).
2.0 BACKGROUND
On January 27, 2006 (ML060900250), FPL submitted an LAR, Spent Fuel Pool Boraflex Remedy, which proposed to eliminate the need to credit Boraflex neutron absorbing material for reactivity control in the Turkey Point SFPs through the use of analyzed new spent fuel storage patterns and Metamic rack inserts. Metamic neutron absorber material is a metal matrix composite consisting of a matrix of 6061 aluminum alloy reinforced with Type 1 ASTM C-750 boron carbide. On July 17, 2007 (ML071800198), the NRC staff issued license amendments related to the Spent Fuel Pool Boraflex Remedy in which reactivity control is performed by a combination of RCCAs, Metamic rack inserts, open water holes, and administrative controls that require mixing higher reactivity fuel with lower reactivity fuel. FPL completed full implementation of the Spent Fuel Pool Boraflex Remedy amendments in 2010 for Turkey Point.
On August 5, 2010 (ML102220022), FPL submitted an LAR, Fuel Storage Criticality Analysis, to revise TS 5.5.1, Fuel Storage-Criticality, to include new spent fuel storage patterns that account for both the increase in fuel maximum enrichment from 4.5 weight percent (wt%)
uranium-235 (U-235) to 5.0 wt% U-235 and the impact on the fuel of higher power operation proposed under the EPU application that was then undergoing review by the NRC staff. The proposed TS changes were based on the results of a new criticality analysis provided as to the letter dated August 5, 2010, WCAP-17094-P, Revision 2, Turkey Point Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis, dated July 2010 (Proprietary).
On February 22, 2011, FPL supplemented its letter of August 5, 2010, with Revision 3 of WCAP-17094-P (ML110560338 - Package) to address then-draft Interim Staff Guidance (ISG)
DSS-ISG-2010-01, Draft Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools. DSS-ISG-2010-01 provides updated guidance to the NRC staff reviewer to address the increased complexity of recent SFP license application analyses and operations.
The draft of DSS-ISG-2010-01 was issued on September 27, 2010, for public comment, and the final ISG was issued on September 29, 2011 (ML110620086). On October 31, 2011 (ML11216A057), the NRC staff issued license amendments that approved the Fuel Storage Criticality Analysis LAR. On February 21, 2012 (ML11293A359 - Package), the NRC staff issued license amendments that approved the Turkey Point EPU.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION On October 11, 2023, FPL submitted the current LAR, involving a new fresh and spent fuel storage criticality analysis in support of revising TS 3.7.13, TS 3.7.14, and TS 4.3 to allow for an updated SFP criticality safety analysis that accounts for the impact on the spent fuel from a proposed transition to 24-month fuel cycles, which is currently under separate review by the NRC staff. The proposed TS changes are based on the results of a new criticality analysis provided as Reference 1 to the October 11, 2023, letter, WCAP-18830-P, Revision 0. The new criticality analysis is heavily based on and represents an evolutionary update to the previous analysis approach presented in WCAP-17094-P, Revision 3. Although fuel storage has been analyzed with respect to 24-month fuel cycles in the new criticality analysis, the transition to 24-month fuel cycles would not be implemented with the issuance of this license amendment.
3.0 REGULATORY EVALUATION
Section 182a of the Atomic Energy Act of 1954, as amended, requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commissions regulatory requirements related to the content of the TSs are contained in Title 10 of the Code of Federal Regulations (10 CFR) 50.36, Technical specifications. The TS requirements in 10 CFR 50.36 include items in 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),
[d]esign 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].
The applicable regulatory requirements for criticality safety analysis for SFPs are contained in 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 62, Prevention of criticality in fuel storage and handling, and in 10 CFR 50.68, Criticality accident requirements.
The regulations in 10 CFR Part 50, Appendix A, Criterion 62 state that:
Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
The regulations in 10 CFR 50.68(b)(2) state, in part, that:
The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The regulations in 10 CFR 50.68(b)(3) state, in part, that:
If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level.
The regulations in 10 CFR 50.68(b)(4) state, in part, that:
If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.
In March of 2021, the NRC staff issued Regulatory Guide (RG) 1.240, Fresh and Spent Fuel Pool Criticality Analyses (ML20356A127). RG 1.240 describes an approach that the NRC staff considers acceptable to demonstrate that applicable regulatory requirements are met for the subcriticality of fuel assemblies stored in fresh fuel vaults and SFPs at light-water reactor power plants. RG 1.240 also endorses, with clarifications and exceptions, the Nuclear Energy Institute (NEI) guidance document NEI 12-16, Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants (ML19269E069). NEI 12-16, Revision 4, is a comprehensive guide that compiles previously issued NRC guidance and clarification letters regarding performing SFP and fresh fuel storage criticality analyses. The NRC staff primarily uses RG 1.240 in conjunction with NEI 12-16, Revision 4, in its review.
In January of 2001, the NRC staff issued NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology (ML050250061). NUREG/CR-6698 provides guidance on the validation and bias and uncertainty quantification of calculational tools used to perform nuclear criticality safety analyses. Guidance is also provided for establishing upper safety limits on subcriticality. The NRC staff primarily used this document to evaluate the SCALE 6.2.4 code package validation and uncertainty quantification contained in Appendix A of WCAP-18830-P, Revision 0, and determined the acceptability of applying SCALE 6.2.4 in the criticality analysis of the Turkey Point SFPs.
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 (ML070570006). Section 9.1.1 provides the existing recommendations for performing the review of the nuclear criticality safety analysis of SFPs.
4.0 TECHNICAL EVALUATION
4.1 Summary In order for the LAR to be acceptable, the licensee must submit a plant-specific SFP criticality analysis that includes technically supported margins. The NRC staff reviewed the analysis to ensure that the assumptions made are technically substantiated. The NRC staff reviewed the application and supplemental information to determine whether the submittals provide
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION reasonable assurance that the regulatory requirements will continue to be met. As discussed below, the NRC staff finds that the licensee provided the technical information needed for the NRC staff to complete its review of the LAR.
4.2 Selection of Design Basis Assembly Section 4 of WCAP-18830-P, Analytical Methodology, indicates that only two fuel assembly designs have been used at Turkey Point: the 15 x 15 Westinghouse Standard (STD) and the 15 x 15 Optimized Fuel Assembly (OFA). Other than grid-type (which is not included in the fuel assembly modeling in the criticality analysis) and structural material composition, the only differences between the fuel assembly designs are the slightly differing inner and outer diameters of the guide tubes and instrument tubes. As a result, the STD fuel assembly design was chosen by the licensee as the design basis fuel assembly design used in the depletion and the SFP environments of the criticality analysis. Section 4.1 and Tables 4-1 through 4-6 of WCAP-17094-P, Revision 3, provide information that supports the selection of STD as the design basis assembly and the associated axial burnup profiles. The supportive documentation in WCAP-17094-P, Revision 3, was previously reviewed by the NRC staff and found to be acceptable.
The design basis assembly is split into three criticality fuel designs, each of which is defined to be representative of the types of fuel assemblies utilized during the most recent core operational phases of Turkey Point: Criticality Fuel Design 1 for unblanketed pre-EPU fuel, Criticality Fuel Design 2 for both pre-EPU and post-EPU mid-enriched blanket design fuel, and Criticality Fuel Design 3 for post-EPU fully enriched fuel.
Since the licensee demonstrated that the STD fuel assembly design is limiting and used it to develop three criticality fuel designs that are each representative of the fuel assemblies utilized in the most recent and expected future core designs, the NRC staff concludes that the method used to determine the limiting design basis assemblies follows the guidance set forth in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240 and, therefore, is acceptable.
4.3 Depletion Analysis Section 4.1 of WCAP-18830-P provides information on the methods used to determine the conservative and bounding inputs for the generation of isotopic number densities. The isotopic number densities are used in the subsequent Monte Carlo simulations of the criticality analyses.
The methodology for depleting fuel assemblies in-reactor to support burnup credit in SFP criticality safety calculations includes the depletion of two-dimensional unit assemblies as an infinite array in the reactor core geometry. This was done using the Westinghouse two-dimensional neutron transport lattice code PARAGON Version 1.4.3 (ML042250345/
ML042250322 - Proprietary/Non-Proprietary, repectively). The depletion analyses were performed using several bounding reactor parameters and a 5-percent reactivity decrement to account for depletion uncertainty, which is consistent with the guidance in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240. The reactor parameters in the analyses include moderator temperature, fuel temperature, soluble boron concentration, axial relative power, and specific power and operating history.
To account for the different operational characteristics of the recent and expected core designs, the isotopic number densities were differentiated by the criticality fuel design, fuel enrichment, burnup, and decay time after discharge. Once depleted to a desired burnup, the fuel was
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION decayed to its most reactive state (100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) and residual Xenon was removed. Criticality Fuel Design 1 had isotopic number densities calculated at 2, 3, and 4 wt% U-235, and decay times of 10, 15, and 20 years were used. Criticality Fuel Designs 2 and 3 had isotopic number densities calculated at 2, 3, 4, and 5 wt% U-235, and decay times of 2.5, 5, 10, 15, 20, and 25 years were used.
To account for the variation in axial relative power, which is consistent with the guidance in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240, the licensee used different approaches for Criticality Fuel Design 1 and Criticality Fuel Designs 2 and 3. For Criticality Fuel Design 1, 1,361 axial burnup profiles for discharge non-blanketed fuel from Unit Nos. 3 and 4 were examined. Data was only available from 12 equally spaced axial nodes, and the most limiting of 3 subsections of nodes was used to create candidate assembly profiles, which are the collection of burnup profiles from which the most limiting burnup profile will be determined. From these candidate assembly profiles, distinct profiles conservative in terms of reactivity were identified for various ranges of burnups. The limiting relative burnup profiles and their associated burnup ranges are listed in Table 4-6 of WCAP-18830-P.
For Criticality Fuel Designs 2 and 3, axial relative burnup profiles for 26 nodes were derived from the operation of post-EPU cycles and the planned operation of 24-month cycles. These profiles were sorted into average burnup bins based on each cycle of reactor operations and in each bin an average nodal burnup was calculated to create a candidate assembly profile.
Assembly profile ranking was performed by comparing four different homogenous combinations of nodes, and from these a minimum nodal burnup profile was chosen. These axial profiles became the limiting burnup profiles and are listed in Tables 4-7 and 4-8 of WCAP-18830-P.
Given the limited number of post-EPU cycles with fully enriched blankets, a further conservatism was introduced by the licensee for these profiles in the form of nodal multipliers that reduce the relative burnup values at each axial elevation. This will effectively increase the reactivity of these assemblies in the subsequent criticality analyses. The NRC staff finds that this is conservative. The nodal multipliers are given in Table 4-3 of WCAP-18830-P.
Since it is typical for fuel modeled with a uniform burnup profile to be more reactive early in life, isotopics for Criticality Fuel Designs 2 and 3 were created and compared using both a uniform burnup profile and a distributed burnup profile. The distributed burnup profile used was the limiting burnup profile for the respective criticality fuel design. The resulting two sets of isotopics were used in the subsequent Monte Carlo criticality calculations to ensure that the most limiting burnup profile (uniform versus distributed) was used to derive the minimum burnup requirements for the various fuel categories used in the SFP storage configurations.
The depletion parameters used to generate the depleted isotopics for Criticality Fuel Design 1 are given in Table 4-4 of WCAP-18830-P. The depletion parameters used to generate the depleted isotopics for Criticality Fuel Designs 2 and 3 are given in Table 4-5 of WCAP-18830-P.
Table 4-9 and Table 4-10 of WCAP-18830-P contain the limiting axial moderator temperature profiles modeled with Criticality Fuel Designs 2 and 3, respectively. Because the licensee either identified the most limiting burnup profile (Criticality Fuel Design 1) or generated a conservative burnup profile (Criticality Fuel Designs 2 and 3), the depleted isotopics for each criticality fuel design will produce in the criticality analysis a more reactive assembly for a given burnup than would be expected from a more representative burnup profile. The NRC staff finds that this is bounding and conservative, and therefore, the NRC staff finds that the licensees analysis of
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION axial burnup profiles is consistent with the guidance in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240.
The depletion analyses included different burnable absorbers depending on the criticality fuel design. Criticality Fuel Design 1 contained PYREX or Westinghouse Wet Annular Burnable Absorber (WABA). Criticality Fuel Design 2 contained WABA during pre-EPU operations and may contain IFBA and Gadolinium absorbers. Criticality Fuel Design 3 may contain IFBA and Gadolinium absorbers. Gadolinium was ignored in the analyses, consistent with guidance in NEI 12-16, Revision 4.
With regard to WABA, Section 4.1.2.2.6, Burnable Absorber Usage, of WCAP-18830-P indicates that, although Criticality Fuel Design 2 contains WABA, isotopics for Criticality Fuel Design 2 were derived for post-EPU conditions using the maximum expected IFBA impact and excluding WABA. This modeling choice was made because it is assumed that all the post-EPU input with IFBA for the analysis is bounding of pre-EPU operation with WABA. However, no justification was provided to support this assumption.
In its October 22, 2024, response to NRC staff RAI-1, the licensee indicated that an analysis was performed to ensure that all assemblies with WABA that were operated in pre-EPU conditions are bounded by Criticality Fuel Design 2 isotopics derived at post-EPU conditions with IFBA. This analysis included a survey of all assemblies containing WABA operated at pre-EPU conditions, a sorting of all those assemblies into similar groups, and comparisons of the depletion reactivity of those groups to that of Criticality Fuel Design 2. The NRC staff examined the results of this analysis, as provided in the response to RAI-1, and determined that the Criticality Fuel Design 2 reactivities at all depletion steps, and therefore the associated isotopics, acceptably bounded or were representative of assemblies containing WABA operated under pre-EPU conditions.
During its review of this issue, the NRC staff noted ((
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
)). Therefore, the NRC staff finds that the modeling choice of deriving Criticality Fuel Design 2 isotopics from post-EPU conditions with maximum IFBA impact and without the consideration of WABA is reasonable.
Hafnium Vessel Flux Depression (HVFD) absorbers are used in a few highly burned fuel assemblies on the core periphery during the third cycle of operation. These absorber inserts are present only near the mid-plane of the fuel assemblys axial length to reduce the fluence at critical weld locations along the core vessel. These absorbers have not been in use in Turkey Point, Unit Nos. 3 and 4, since Cycles 23 and 24, respectively, and they are not used in post-EPU fuel. For all assemblies that contained hafnium inserts, the distributed burnup profile was limiting, since the HVFDs push the neutron flux out of the middle of the core and toward the ends of the fuel during fuel depletion. This elevated flux results in an over depletion of the ends of the fuel and results in an assembly with hafnium inserts being bounded by the same fuel assembly that is depleted without HVFD absorbers.
Since the effect of significant rod insertion (greater than 20 centimeters (cm)) can affect the axial burnup profile, the licensee stated that Turkey Point does not operate at full power with significant amounts of rod insertion. If any assemblies are operated with significant rod insertion, they must not credit rodded burnup. Because the burnup accrued while under rodded operation is not credited in the depletion analysis, the assembly reactivity in the subsequent criticality analysis will be greater than it would otherwise be, which is conservative. Therefore, the NRC staff finds that this approach is reasonable.
Since the licensee accounted for reactivity uncertainty with a 5-percent decrement, performed depletion simulations with reactor parameters that maximize the reactivity of the depleted fuel assembly, and appropriately considered burnable absorbers and rodded operations, the NRC staff concludes that the depletion analyses are consistent with the guidance set forth in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240. The NRC staff therefore finds that the depletion analyses are acceptable.
4.4 Criticality Analysis Section 4.2 of WCAP-18830-P provides information on criticality analysis and the reactivity calculations and evaluations used in developing the burnup requirements for the Turkey Point SFPs. Models for KENO, the Monte Carlo module in the SCALE 6.2.4 code package, were generated for all criticality fuel designs and for the proposed storage arrays within Region I and Region II of the SFPs. The Cask Area Rack was also modeled. Assembly initial enrichment, average burnup, and decay time (or fresh IFBA) were varied to determine storage limits.
Assembly storage is controlled by assigning fuel categories to each storage location in a storage configuration. Multiple storage configurations are defined, each a 2 x 2 array of fuel storage locations. Table 4-11, Fuel Categories Rank by Reactivity, of WCAP-18830-P provides the fuel categories ranked by reactivity and the regions where they would be placed in the SFP. In all configurations, an assembly of lower reactivity can replace an assembly of higher reactivity, or the cell can be left empty (water-filled). Figures 4-1 and 4-2 of WCAP-18830-P describe the allowed configurations of fuel categories in the SFP. The removable Cask Area Rack is modeled for unrestricted storage as it is modeled as a flux-trap style rack with Boral absorber inserts (and a requirement that the boral areal density remains above ((
))). Cask Area Rack and Fuel Rod Basket information is covered in more detail below.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION For assembly storage by fuel configuration, minimum burnup limits were determined using the depletion analyses. With these minimum burnup limits, the equation in Section 5.2.1 of WCAP-18830-P was used along with Tables 5-2 through 5-4 of WCAP-18830-P to determine the burnup requirements for fuel as a function of initial enrichment and decay time. Fuel categories I-1 and I-2 do not have minimum burnup requirements due to being fresh unburned fuel.
As provided in 10 CFR 50.68(b)(4), when credit is taken for soluble boron, the k-effective (k-eff) of the SFP storage racks must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. To calculate the target k-eff, the maximum allowable k-eff for a 2 x 2 fuel storage array was subtracted by the administrative margin of 0.005 k and the summation of the reactivity that accounts for biases and uncertainties in the reactivity calculations. Uncertainties account for variations such as manufacturing tolerances, analytical uncertainties, or measurement uncertainties. Except for the case of the ((
)). Uncertainties were added as the root sum square of the individual reactivity uncertainties. Tables 4-12 through 4-36 of WCAP-18830-P contain the biases, uncertainties, and the target k-eff (when including the administrative margin) for each defined 2 x 2 fuel storage array. These results were determined by analyzing each defined 2 x 2 fuel storage array in an infinite, repeating pattern. The NRC staff notes that the k-eff results for all analyzed 2 x 2 fuel storage arrays are less than 1.0, which meets the applicable regulatory requirements for the SFP storage racks.
The interface of fuel storage arrays within Region I or Region II must adhere to certain requirements. The primary interface requirement identified in WCAP-18830-P is that each 2 x 2 fuel storage array within a region must match one of the analyzed arrays. In this context, match is defined as each array must have the required number of inserts (e.g., full length RCCAs or empty cells), and that the assemblies must have at least the required burnup for the appropriate category. Following these requirements ensures that every 2 x 2 fuel storage array interface within a region matches an analyzed condition. And, as discussed above, each of these analyzed conditions meets the applicable k-eff regulatory requirements for the SFP storage racks.
The 2 x 2 storage arrays at the Region I to Region II interface require restrictions in addition to those imposed on the intra-region interfaces. Analyses performed by the licensee identify allowable 2 x 2 fuel storage arrays at the Region I to Region II interface. Each 2 x 2 fuel storage array at the Region I to Region II interface must match one of the analyzed configurations.
Following these requirements, every 2 x 2 configuration across regional interfaces will correspond to an analyzed condition. The analyzed configurations are illustrated in Figure 5-1, Figure 5-2, and Figure 5-3 of WCAP-18830-P, and Table 5-31 lists the analyzed configurations and the associated k-eff results. The NRC staff notes that the k-eff results including all biases and uncertainties for all analyzed 2 x 2 arrays are less than 1.0 and, therefore, meet the applicable k-eff regulatory requirements.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The fuel rod basket is used to store loose fuel rods, as needed. To analyze the effects that it may have on reactivity, it was modeled as if every tube in the basket was filled with fresh 5.0 wt% U-235 fuel rods. Due to inserts being mounted on top of the assembly, it is not possible to place a fuel rod basket in a location requiring an insert. The fuel rod basket was modeled in both Regions I and II with fresh and burned fuel. The results given in Table 5-29 and Table 5-30 of WCAP-18830-P show that replacing an analyzed assembly with a fuel rod basket results in a reactivity decrease for all configurations. Therefore, a fuel rod basket is acceptable in any cell without Metamic inserts that is not designated as an empty location for the storage configuration.
Regarding the Cask Area Rack, sufficient absorber panels are present that the maximum k-eff is much less than the limiting k-eff in Region I or Region II. Therefore, there were no interface loading constraints on the Cask Area Rack to Region I or II interfaces.
New fuel storage racks must meet the appropriate acceptance criteria per 10 CFR 50.68(b)(2) and (b)(3). Specifically, if the fresh fuel storage racks are flooded with full density unborated water with fuel of maximum fuel reactivity, the k-eff shall not exceed 0.95, at a 95 percent probability and 95 percent confidence level. If flooded with a low-density hydrogenous fluid with optimum moderation conditions and fuel of maximum fuel assembly reactivity, the k-eff shall not exceed 0.98, at a 95 percent probability and 95 percent confidence level. Each criterion must be met including all biases and uncertainties. The NRC staff noted that the licensees analyses demonstrate that these conditions are met for all fuel enriched to 4.25 wt% U-235 or less. The licensees analyses also show that 5.0 wt% U-235 fuel will meet the respective fully flooded and optimum moderation k-eff acceptance criteria of 0.95 and 0.98 in the fresh fuel rack without additional absorbers when 16 IFBA rods are loaded in the assembly, which is the minimum number of rods in any Westinghouse IFBA design. These designs contain IFBAs that cover more than the ((
)).
Section 5.5, Other Normal Storage Conditions, of WCAP-18830-P provides information on normal SFP activities that occur in addition to normal fuel handling and storage (e.g.,
reconstitution of fuel assembly, ultrasonic testing of fuel assembly, fuel assembly inspection, etc.). The licensee listed 15 normal conditions that were taken into consideration in the criticality analyses along with justifications for why they are acceptably bounded. The NRC staff reviewed these justifications and determined that they are reasonable because the additional activities either do not impact the normal fuel handling and storage analyses or the additional activities result in decreased reactivity of the stored fuel assembly. Therefore, the requirement of 10 CFR 50.68(b)(4) that the k-eff shall not exceed 0.95 is met for these activities.
Section 5.7.2, Soluble Boron for Accident Conditions, of WCAP-18830-P and the licensees response to NRC staff RAI-2 provide information on accident conditions. The licensee listed the accident conditions that were taken into consideration in the analysis, which are: misloaded fresh fuel assembly or assemblies in a storage rack; inadvertent removal of an absorber insert; SFP 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; misplacement of cask area rack (assembly in corner); and misplacement of cask area rack (mis-rotated 180°). All accident calculations were performed with 2300 parts per million (ppm) of soluble boron. Consistent with the guidance in NEI 12-16, Revision 4, the licensee included an additional 50 ppm margin to account for the reactivity effects of not calculating biases and uncertainties in borated conditions and not modeling spacer grids. The NRC staff notes that this meets the proposed TS limit of 2350 ppm of soluble boron.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The licensee stated that the ((
)) was bounding of all other accident scenarios. The ((
)). In the case of the ((
)), after accounting for the biases and uncertainties in borated conditions, the licensee determined an administrative margin of 0.0025 k was acceptable to meet regulatory requirements.
To show that the ((
)) was bounding, the licensee provided a list of justifications in its October 22, 2024, response to NRC staff RAI-2 along with the results of analyses. The results of the analyses can be found in Table 4-1 below, with the exception of the dropped fuel assembly and the misplaced fuel assembly cases. The licensee asserted that the dropped fuel assembly accident does not need to be analyzed because the separation between the dropped fuel assembly and remaining fuel due to the top nozzle, fuel rod plenum, fuel rod end plugs, and the separation between the fuel rod and top nozzle is greater than 12 inches; therefore, only a small positive reactivity change would occur, which is consistent with NEI 12-16, Revision 4. For the misplaced fuel assembly, the licensee concluded that these conditions are bounded by the misload event because any fuel assembly placed outside of the racks is surrounded by water on at least two sides and encounters additional neutron leakage as opposed to the ((
)). The NRC staff reviewed the justifications and, with consideration of prior SFP criticality analysis reviews (e.g., ML11293A366 /
ML11293A365 - Proprietary/Non-Proprietary, repectively), found them to be reasonable.
Table 4-1: Results of the Accident Calculations Accident Description Max k-eff + 2 sigma Infinite Model of Fresh 5 wt% U-235 Assemblies
((
))
Misloaded fresh fuel assembly or assemblies in a storage rack ((
))
((
))
Misloaded fresh fuel assembly or assemblies in a storage rack ((
)) including biases and uncertainties*
((
))
Inadvertent removal of absorber inserts
((
))
Spent fuel temperature greater than normal operating range
((
))
Loss of water gap between Region I and Region II due to seismic event
((
))
Dropped fresh fuel assembly N/A Misplaced fuel assembly N/A Misplacement of Cask Area Rack (assembly in corner)
((
))
Misplacement of Cask Area Rack (mis-rotated 180°)
((
))
- This is the only case to use biases and uncertainties since it is bounding of all other accident scenarios.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION In all cases, accident condition analyses provided a k-eff that is within the regulatory requirement of 0.95 specified in 10 CFR 50.68(b)(4) for when soluble boron is credited.
Therefore, based on this and the discussions above, the NRC staff finds that the normal activities and the accident condition analyses are acceptable.
4.5 Criticality Code Validation The criticality analysis methodology uses the Westinghouse two-dimensional neutron transport lattice code PARAGON Version 1.4.3, as well as SCALE Version 6.2.4. In the present application, PARAGON is used for simulation of in-reactor fuel assembly depletion to generate fuel isotopic concentrations for burnup credit and SCALE is used for reactivity determinations of fuel assemblies in the SFP conditions. PARAGON is generically approved by the NRC for depletion calculations (ML042250322). Validation of SCALE for the present application is provided in Appendix A of WCAP-18830-P and utilizes a combination of criticality benchmark experiments (from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and NUREG/CR-6361 (ML21042B285)) and the Haut Taux de Combustion (HTC) critical experiment data. The validation is intended to cover fresh and spent fuel storage for Turkey Point, including the criticality analysis of all normal operations and postulated accidents in the SFPs and fresh fuel storage.
In validating SCALE for the Turkey Point SFP criticality safety analysis, Westinghouse applied guidance from NUREG/CR-6698 and NUREG-1475, Revision 1, Applying Statistics (ML11102A076). Per NUREG/CR-6698, key physical parameters to be considered when defining the area of applicability of a benchmark suite fall into three categories: materials, geometry, and neutron energy spectrum. Key physical parameters within these three categories may be broadly delineated as:
- 1) Fissile isotope
- 2) Enrichment of fissile isotope
- 3) Fuel density
- 4) Fuel chemical form
- 5) Type of neutron moderators and reflectors
- 6) Range of moderator to fissile isotope
- 7) Neutron absorbers
- 8) Physical configurations The NRC staff notes that additional constituent parameters may be necessary to fully characterize the parameters listed above (e.g., reflector density and reflector isotopic composition). The full list of recommended physical parameters may be found in Table 2.3 of NUREG/CR-6698. Table A-21 of WCAP-18830-P summarizes the key physical parameters for the current validation and identifies the ranges of the parameters supported by the selected critical experiments. The NRC staff examined the selected critical experiments included in the validation suite (particularly with respect to varying moderator temperature, concrete reflectors, and fissionable material isotopic composition) and determined, with one exception, that they adequately cover fresh and spent fuel storage for the present and anticipated non-mixed-oxide light water reactor fuel designs at Turkey Point with respect to the list of key physical parameters discussed above.
The single exception pertains to the neutron absorber material type and physical configuration for ((
)). The selected critical experiments do not
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION consider ((
)), and thus the accuracy with which SCALE can model this is unknown. The guidance provided in Section A.1.1 of NEI 12-16, Revision 4, for the range of parameters to be validated for criticality codes states, with regard to isotopic content, [e]xperiments should cover materials representative of the rack structure and others if used in the criticality analysis.
Further, the guidance provided in Section 2.2 of NUREG/CR-6698 states, the critical experiments selected for inclusion in the validation must be representative of the types of materials, conditions, and operating parameters found in the actual operations to be modeled. As indicated in Section 4.2.1 of WCAP-18830-P, the ((
)) is necessary to maintain multiple fresh and SFP storage arrays subcritical and within acceptable regulatory limits in unborated conditions.
NEI 12-16, Revision 4, indicates that the influence that ((
)) have on neutronic behavior should be appropriately captured and considered with respect to quantified biases and uncertainties. Consistent with this, Section 4.2.3 of WCAP-18830-P identifies that ((
)) are conservatively modeled by ((
)). While the NRC staff finds that this approach introduces conservatism, it is not clear from the submittal what the approximate magnitude is of the introduced conservatism. It is possible that the conservatism may not be sufficient to offset a potential additional source of uncertainty, that which might stem from the SCALE modeling of ((
)), and the quantified bias and uncertainty reported in WCAP-18830-P for SCALEs ability to model fresh and/or spent fuel with neutron absorbers (discussed further below) would not be appropriate for determining margin to regulatory limits on k-eff.
In its October 22, 2024, response to NRC staff RAI-3, the licensee identified that the approximate minimum magnitude of ((
)) modeling conservatism is ((
)), and the NRC staff notes that the average conservatism of the results provided in this response is approximately ((
)). Table A-19 of WCAP-18830-P indicates that the quantified bias uncertainties for ((
)). While the accuracy with which SCALE can model the neutronic influence of ((
)) is unknown, the existing validation suite includes an array of neutron absorbing materials, and the influence of these materials are quantified in the reported bias uncertainties.
As such, the NRC staff anticipates that the resulting increase in the bias uncertainty, if any, due to the inclusion of an additional neutron absorber material (((
))) would be minor, between 5 and 10 percent. By comparison, the minimum modeling conservatism of ((
))
can offset at least a 17-percent increase in the bias uncertainties. Additionally, the average modeling conservatism of ((
)) can offset at least a 50-percent increase in the bias uncertainty. An increase in bias uncertainty of this magnitude is unrealistic and would suggest that a more serious underlying issue exists with the application of the SCALE code (e.g.,
incorrect modeling geometry, incorrect or corrupted nuclear data files, etc.). Such an issue would already be apparent from the supplied analytical results. Therefore, the NRC staff finds that the conservative modeling approach is sufficient to offset a possible increase in the uncertainty associated with modeling fresh and/or spent fuel with neutron absorbers. The NRC staff also finds that the influence of ((
)) on neutronic behavior has been appropriately accounted for and considered with respect to biases and uncertainties, consistent with the guidance in NEI 12-16, Revision 4, and that its omission from the validation suite is acceptable. Based on this and the discussion
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION above, the NRC staff finds that the critical experiments selected for the validation of SCALE are consistent with the guidance provided in NUREG/CR-6698 and are, therefore, acceptable.
Regarding uncertainty quantification, Westinghouse applied the guidance in NUREG/CR-6698 to determine the SCALE methodology bias and bias uncertainty. The SCALE methodology bias and bias uncertainty is used in combination with other biases and uncertainties (e.g.,
manufacturing uncertainties) and additional subcritical margin to establish an upper safety limit that ultimately ensures that regulatory requirements are met.
Several different types of calculations frequently occur with respect to analyzing fuel storage.
These calculations ((
)).
The NRC staff finds this approach reasonable because it is representative of the broad categorizations of the cells in each analyzed fresh and SFP storage array when a fuel assembly is present.
The raw calculational results were subject to goodness-of-fit tests for normality to discern the appropriate statistical treatment to use when analyzing the data (e.g., normally distributed or non-parametric). Trending analyses, consisting of regression fits to calculated results, were also performed, when applicable, to further inform the appropriate approach to determining the bias and uncertainty of each dataset and the associated tolerance limit (e.g., one-sided tolerance limit or tolerance band). The trending analyses were applied to each subset of normally distributed data for multiple key physical parameters. ((
)). Consistent with the guidance provided in NUREG/CR-6698, ((
)). The NRC staff finds this approach reasonable because it introduces additional conservatism.
Based on the goodness-of-fit tests and the trending analyses, the NRC staff concludes that appropriate statistical treatments consistent with NUREG/CR-6698 and NUREG-1475, Revision 1, were applied, and that biases and uncertainties were appropriately determined ((
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION
)). The NRC staff also concludes that the validation of SCALE to perform criticality safety calculations was performed in accordance with the guidance in NUREG/CR-6698 and NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240. Therefore, the NRC staff finds that SCALE Version 6.2.4 is acceptable for the present application.
4.6 Technical Conclusion The licensee has demonstrated through its submittal, as supplemented, that the methodologies used in its criticality analysis follow the guidance set forth in NEI 12-16, Revision 4, as endorsed by the NRC in RG 1.240, as well as other applicable guidance. Therefore, the NRC staff finds that the proposed TS changes that are consistent with this criticality analysis comply with the applicable regulatory requirements of 10 CFR 50.36, 10 CFR Part 50, Appendix A, and 10 CFR 50.68 and are acceptable.
5.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Florida State official was notified of the proposed issuance of the amendments on November 12, 2024. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on January 23, 2024 (89 FR 4342) and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: K. Heller, NRR J. Vande Polder, NRR Date: December 11, 2024