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| issue date = 09/19/2012
| issue date = 09/19/2012
| title = Public - Issuance of Amendment Regarding New Fuel Vault and Spent Fuel Pool Nuclear Criticality Analysis
| title = Public - Issuance of Amendment Regarding New Fuel Vault and Spent Fuel Pool Nuclear Criticality Analysis
| author name = Orf T J
| author name = Orf T
| author affiliation = NRC/NRR/DORL/LPLII-2
| author affiliation = NRC/NRR/DORL/LPLII-2
| addressee name = Nazar M
| addressee name = Nazar M
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 September 19, 2012 Mr. Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420 ST. LUCIE PLANT, UNIT 2 -ISSUANCE OF AMENDMENT REGARDING NEW FUEL VAULT AND SPENT FUEL POOL NUCLEAR CRITICALITY ANALYSIS (TAC NO. ME8782)  
{{#Wiki_filter:OFFICIAL USE ONLY         PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 19, 2012 Mr. Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420
 
==SUBJECT:==
ST. LUCIE PLANT, UNIT 2 - ISSUANCE OF AMENDMENT REGARDING NEW FUEL VAULT AND SPENT FUEL POOL NUCLEAR CRITICALITY ANALYSIS (TAC NO. ME8782)


==Dear Mr. Nazar:==
==Dear Mr. Nazar:==
The Commission has issued the enclosed Amendment No. 162 to Renewed Facility Operating License No. NPF-16 for the st. Lucie Plant, Unit No.2. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 25, 2011, as supplemented by letters dated November 4 and December 8, 2011, and April 30 and May 4 and 7, 2012. This amendment raises the maximum fuel enrichment for fresh fuel storage from a maximum of 4.5 weight-percent uranium-235 to a maximum lattice averaged value of 4.6 weight-percent uranium-235.
 
The TS changes associated with fuel stored in the spent fuel pool include increasing the maximum initial enrichment from 4.5 weight-percent uranium-235 to a maximum planar average initial enrichment of 4.6 weight-percent uranium-235, credit for empty storage locations, credit for use of METAMICTM inserts, credit for installation of full-length full-strength, five-fingered control element assemblies, and definition of three special configurations referred to in the nuclear criticality safety analysis as inspection and maintenance configurations.
The Commission has issued the enclosed Amendment No. 162 to Renewed Facility Operating License No. NPF-16 for the st. Lucie Plant, Unit No.2. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 25, 2011, as supplemented by letters dated November 4 and December 8, 2011, and April 30 and May 4 and 7, 2012.
OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY M. Nazar -2 The NRC has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations, Section 2.390, "Public Inspections, Exemptions, Requests for Withholding." Accordingly, the NRC staff has also prepared a redacted, publicly-available, non-proprietary version of the SE. Copies of theproprietary and non-proprietary versions of the SE are enclosed.
This amendment raises the maximum fuel enrichment for fresh fuel storage from a maximum of 4.5 weight-percent uranium-235 to a maximum lattice averaged value of 4.6 weight-percent uranium-235. The TS changes associated with fuel stored in the spent fuel pool include increasing the maximum initial enrichment from 4.5 weight-percent uranium-235 to a maximum planar average initial enrichment of 4.6 weight-percent uranium-235, credit for empty storage locations, credit for use of METAMICTM inserts, credit for installation of full-length full-strength, five-fingered control element assemblies, and definition of three special configurations referred to in the nuclear criticality safety analysis as inspection and maintenance configurations.
A copy of the Safety Evaluation is also enclosed.
OFFICIAL USE ONLY         PROPRIETARY INFORMATION
The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, Tracy J. Ort, Project Manager Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389  
 
OFFICIAL USE ONLY         PROPRIETARY INFORM,AJ"ION M. Nazar                                           -2 The NRC has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations, Section 2.390, "Public Inspections, Exemptions, Requests for Withholding." Accordingly, the NRC staff has also prepared a redacted, publicly-available, non-proprietary version of the SE. Copies of theproprietary and non-proprietary versions of the SE are enclosed.
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely, Tracy J. Ort, Project Manager Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389


==Enclosures:==
==Enclosures:==
: 1. Amendment No. 162 to NPF-16 2. Non-Proprietary Safety Evaluation
: 1. Amendment No. 162 to NPF-16
: 3. Proprietary Safety Evaluation cc wI Enclosures 1 and 2: Distribution via Listserv OFFICIAL USE ONLY PROPRIETARY UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 FLORIDA POWER & LIGHT ORLANDO UTILITIES COMMISSION THE CITY OF ORLANDO, FLORIDA MUNICIPAL POWER DOCKET NO. ST. LUCIE PLANT UNIT AMENDMENT TO RENEWED FACILITY OPERATING Amendment No. 162 Renewed License No. NPF-16 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by Florida Power & Light Company (the licensee), dated February 25,2011, as supplemented by letters dated November 4 and December 8,2011, and April 30 and May 4 and 7, 2012; complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 Commission's regulations; 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 The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
: 2. Non-Proprietary Safety Evaluation
-2 Accordingly, Renewed Facility Operating License No. NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 3.B to read as follows: Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 162, are hereby incorporated in the renewed license. FPL shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of its date of issuance and shall be implemented within 60 days. FOR THE NUCLEAR REGULATORY COMMISSION Ie . Quichocho, Acting Chief ant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation  
: 3. Proprietary Safety Evaluation cc wI Enclosures 1 and 2: Distribution via Listserv OFFICIAL USE ONLY         PROPRIETARY INFORM,A,TION
 
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FLORIDA POWER & LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO, FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST. LUCIE PLANT UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 162 Renewed License No. NPF-16
: 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Florida Power & Light Company (the licensee), dated February 25,2011, as supplemented by letters dated November 4 and December 8,2011, and April 30 and May 4 and 7, 2012; complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.
 
                                                -2
: 2. Accordingly, Renewed Facility Operating License No. NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 3.B to read as follows:
B.      Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 162, are hereby incorporated in the renewed license.
FPL shall operate the facility in accordance with the Technical Specifications.
: 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY COMMISSION
                                                    -'7J7_~'~~
Ie . Quichocho, Acting Chief ant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation


==Attachment:==
==Attachment:==


Changes to the Operating License and Technical Specifications Date of Issuance:
Changes to the Operating License and Technical Specifications Date of Issuance: September 18, 2012
September 18, 2012 ATTACHMENT TO LICENSE AMENDMENT NO. TO RENEWED FACILITY OPERATING LICENSE NO. DOCKET NO. Replace Page 3 of Renewed Operating License NPF-16 with the attached Page 3. Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and
 
ATTACHMENT TO LICENSE AMENDMENT NO. 162 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 Replace Page 3 of Renewed Operating License NPF-16 with the attached Page 3.
Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Pages                              Insert Pages XXII                                      XXII XXV                                      XXV 3/49-12                                  3/49-12 5-4                                      5-4 5-4a                                      5-4a 5-4b                                      5-4b 5-4c 5-4d 5-4e 5-4f 5


==1.0 INTRODUCTION==
TABLE 5.6-1 Minimum Burnup Coefficients Cooling Time                            Coefficients Fuel Type            (Years)                A                  B                  C 1                  0            -33.4237            25.6742              -1.6478 I
2                    0            -25.3198            14.3200            -0.4042 i      3                  0            -23.4150            16.2050            -0.5500 0              -33.2205            24.8136              -1.5199 I                        2.5            -31.4959            23.4776              -1.4358 i                          5              -30.4454            22.7456              -1.4147 4
10            -28.4361            21.2259              -1.2946 I                          15            -27.2971            20.3746              -1.2333 20              -26.1673            19.4753              -1.1403 0              -24.8402            23.5991              -1.2082 2.5            -22.9981            21.6295              -1.0249 5              -21.8161            20.5067            -0.9440 5
10              -20.0864            19.0127            -0.8545 15              -19.4795            18.3741            -0.8318 20              -18.8225            17.7194            -0.7985 0            -32.4963            25.3143            -1.5534 2.5            -30.6688            23.6229            -1.4025 5            -29.2169            22.5424            -1.3274 i        6 10            -27.2539            21.0241            -1.2054 15            -25.7327            19.8655            -1.1091 20            -25.2717            19.5222            -1.1163 0            -24.6989            24.1660            -1.2578 2.5            -23.0399            22.3047            -1.0965 5            -21.2473            20.6553            -0.9403 7
10            -20.1775            19.5506            -0.9015 15            -19.4037            18.6626            -0.8490 20            -18.3326            17.7040            -0.7526 8                  0            -43.4750            11.6250            0.0000 NOTES:
: 1. To qualify in a "fuel type", the burnup of a fuel assembly must exceed the minimum burnup "SU" calculated by inserting the "coefficients" for the associated "fuel type" and "cooling time" into the following polynomial function:
SU  =A + S'*E + C*E2, where:
SU  = Minimum Bumup (GWD/MTU)
E =Maximum Initial Planar Average Enrichment (weight percent U-235)
A, S, C =Coefficients for each fuel type
: 2. Interpolation between values of cooling time is not permitted.
ST. LUCIE* UNIT 2                                  5-40                  Amendment No. 162
 
OFFICIAL USE ONLY        PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 162 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 FLORIDA POWER AND LIGHT COMPANY, ET AL.
ST. LUCIE PLANT. UNIT NO.2 DOCKET NOS. 50-389
 
==1.0     INTRODUCTION==


By letter dated February 25,2011 (Reference 1), Florida Power and Light Company (the licensee) requested to amend Renewed Facility Operating License No. NPF-16 and revise the St. Lucie, Unit No.2 (St. Lucie 2), Technical Specifications (TSs). The proposed amendment requested revisions to the Renewed Facility Operating License and TSs to support an extended power uprate (EPU) for St. Lucie, Unit 2, at a licensed core thermal power level increased from 2700 megawatts thermal (MWt) to 3020 MWt. For scheduling purposes, the review of the nuclear criticality safety (NCS) analysis for the St. Lucie 2 spent fuel pool (SFP) and new fuel vault (NFV) has been processed as a separate amendment request. The following evaluation presents the results of the U.S. Nuclear Regulatory Commission's (NRC) review of the NCS analysis.
By letter dated February 25,2011 (Reference 1), Florida Power and Light Company (the licensee) requested to amend Renewed Facility Operating License No. NPF-16 and revise the St. Lucie, Unit No.2 (St. Lucie 2), Technical Specifications (TSs). The proposed amendment requested revisions to the Renewed Facility Operating License and TSs to support an extended power uprate (EPU) for St. Lucie, Unit 2, at a licensed core thermal power level increased from 2700 megawatts thermal (MWt) to 3020 MWt. For scheduling purposes, the review of the nuclear criticality safety (NCS) analysis for the St. Lucie 2 spent fuel pool (SFP) and new fuel vault (NFV) has been processed as a separate amendment request. The following evaluation presents the results of the U.S. Nuclear Regulatory Commission's (NRC) review of the NCS analysis.
The License Amendment Request (LAR) was amended by a letter dated November 4, 2011 (Reference 2). Attachment 3 to that letter included Revision 2 to Holtec International Report HI-21 04753. In subsequent submittals, the licensee provided revised proposed technical specifications (Reference 3); responses to NRC requests for additional information (RAls) (References 4,5, and 6); Revision 4 to Holtec International Report HI-21 04753 (HI-21 04753) (Attachment 3 to Reference 6); and Holtec International Report HI-2094416, Revision 1, the NCS analysis for the St. Lucie 2 new fuel vault (Attachment 4 to Reference 6). As part of Attachment 5 to Reference 1 and in support of the EPU, the licensee submitted Holtec International Report HI-2104753, Revision 1, "St. Lucie Unit 2 Criticality Analysis for EPU and Non-EPU Fuel," which describes the NCS analysis for the St. Lucie 2 SFP storage racks. It describes the methodology and analytical models used in the NCS analysis to show that the storage racks maximum k-effective (k eff) will be less than 1.0 when flooded with unborated water for normal conditions, and less than or equal to 0.95 when flooded with borated water for normal and credible accident conditions at a 95-percent probability, 95-percent confidence level. OFFICIAL USE ONLY PROPRIET)\RY OFFICIAL USE ONLY PROPRIETARY INFORMATION -2 The St. Lucie 2 SFP has three fuel storage rack types, with one or more racks for each type: one Cask Pit Rack (CPR). 6 Region 1 Spent Fuel Racks (SFRs) that contain full-length nonstructural L-shaped stainless steel inserts that serve to enhance local neutron absorption, and 13 Region 2 SFRs. HI-21 04753, Revision 4 proposes 10 storage configurations, 2 in the CPR, 3 in the Region I SFRs, and 5 in the Region 2 SFRs. The CPR contains and credits empty cells, fuel burnup, and the neutron absorbing material BORALTM. The Region 1 SFRs credit fuel burnup. empty cells, control rods, and steel inserts. The Region 2 SFRs credit higher burnups, empty cells, control rods, cooling time, and inserts made from the neutron absorbing material METAMICTM.
The License Amendment Request (LAR) was amended by a letter dated November 4, 2011 (Reference 2). Attachment 3 to that letter included Revision 2 to Holtec International Report HI-21 04753. In subsequent submittals, the licensee provided revised proposed technical specifications (Reference 3); responses to NRC requests for additional information (RAls)
The presence of soluble boron is credited in the St. Lucie 2 SFP. New fuel assemblies may be stored in what are normally dry conditions in the St. Lucie 2 new fuel racks in the NFV. It is necessary to update the NFV NCS analysis for the EPU because the maximum planar average enrichment has been increased to 4.6 weight-percent uranium-235.
(References 4,5, and 6); Revision 4 to Holtec International Report HI-21 04753 (HI-21 04753)
No credit is taken in fresh or spent fuel storage racks for integral burnable absorbers.  
(Attachment 3 to Reference 6); and Holtec International Report HI-2094416, Revision 1, the NCS analysis for the St. Lucie 2 new fuel vault (Attachment 4 to Reference 6).
As part of Attachment 5 to Reference 1 and in support of the EPU, the licensee submitted Holtec International Report HI-2104753, Revision 1, "St. Lucie Unit 2 Criticality Analysis for EPU and Non-EPU Fuel," which describes the NCS analysis for the St. Lucie 2 SFP storage racks. It describes the methodology and analytical models used in the NCS analysis to show that the storage racks maximum k-effective (keff) will be less than 1.0 when flooded with unborated water for normal conditions, and less than or equal to 0.95 when flooded with borated water for normal and credible accident conditions at a 95-percent probability, 95-percent confidence level.
OFFICIAL USE ONLY       PROPRIET)\RY INFORM)\TION
 
OFFICIAL USE ONLY         PROPRIETARY INFORMATION
                                                -2 The St. Lucie 2 SFP has three fuel storage rack types, with one or more racks for each type:
one Cask Pit Rack (CPR). 6 Region 1 Spent Fuel Racks (SFRs) that contain full-length nonstructural L-shaped stainless steel inserts that serve to enhance local neutron absorption, and 13 Region 2 SFRs. HI-21 04753, Revision 4 proposes 10 storage configurations, 2 in the CPR, 3 in the Region I SFRs, and 5 in the Region 2 SFRs. The CPR contains and credits empty cells, fuel burnup, and the neutron absorbing material BORALTM. The Region 1 SFRs credit fuel burnup. empty cells, control rods, and steel inserts. The Region 2 SFRs credit higher burnups, empty cells, control rods, cooling time, and inserts made from the neutron absorbing material METAMICTM. The presence of soluble boron is credited in the St. Lucie 2 SFP.
New fuel assemblies may be stored in what are normally dry conditions in the St. Lucie 2 new fuel racks in the NFV. It is necessary to update the NFV NCS analysis for the EPU because the maximum planar average enrichment has been increased to 4.6 weight-percent uranium-235.
No credit is taken in fresh or spent fuel storage racks for integral burnable absorbers.
 
==2.0      REGULATORY EVALUATION==
 
Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, Criterion 62 requires, "Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations."
Paragraph 50.68(b)(1) of 10 CFR requires, "Plant procedures shall prohibit the handling and storage at anyone time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water."
Paragraph 50.68(b)(2) of 10 CFR requires, "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. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used."
Paragraph 50.68(b)(3) of 10 CFR requires, "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. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used."
Paragraph 50.68(b)(4) of 10 CFR requires, "If no credit for soluble boron is taken, 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 unborated water. 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."
OFFICIAL USE ONLY        PROPRIETARY INFORMATION


==2.0 REGULATORY EVALUATION==
OFFICIAL USE ONLY      PROPRIETARY INFORM,tHION
                                                -3 Paragraph 50.36(c)(4) of 10 CFR requires, "Design features. 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 this section."
The St. Lucie 2 SFP NCS analysis does take credit for soluble boron for normal operating conditions. Therefore, the regulatory requirement is for the St. Lucie 2 keff 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 keff must remain below 1.0 (subcritical), at a 95-percent probability, 95-percent confidence level, if flooded with unborated water. The St. Lucie 2 NCS uses the double contingency principle to take credit for soluble boron for abnormal/accident operating conditions.
The St. Lucie 2 NFV NCS analysis demonstrated that, consistent with the requirements of 10 CFR 50.68, the maximum keff at full water density was less than 0.95, at a 95-percent probability/95-percent confidence level and that the maximum keff at optimum water density, around 8 percent of full density, was less than 0.98, at a 95-percent probability, 95-percent confidence level.


Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, Criterion 62 requires, "Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations." Paragraph 50.68(b)(1) of 10 CFR requires, "Plant procedures shall prohibit the handling and storage at anyone time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water." Paragraph 50.68(b)(2) of 10 CFR requires, "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. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used." Paragraph 50.68(b)(3) of 10 CFR requires, "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. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used." Paragraph 50.68(b)(4) of 10 CFR requires, "If no credit for soluble boron is taken, 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 unborated water. 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." OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORM,tHION -3 Paragraph 50.36(c)(4) of 10 CFR requires, "Design features.
==3.0     TECHNICAL EVALUATION==
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 this section." The St. Lucie 2 SFP NCS analysis does take credit for soluble boron for normal operating conditions.
Therefore, the regulatory requirement is for the St. Lucie 2 keff 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 keff must remain below 1.0 (subcritical), at a 95-percent probability, 95-percent confidence level, if flooded with unborated water. The St. Lucie 2 NCS uses the double contingency principle to take credit for soluble boron for abnormal/accident operating conditions.
The St. Lucie 2 NFV NCS analysis demonstrated that, consistent with the requirements of 10 CFR 50.68, the maximum keff at full water density was less than 0.95, at a 95-percent probability/95-percent confidence level and that the maximum keff at optimum water density, around 8 percent of full density, was less than 0.98, at a 95-percent probability, 95-percent confidence level. 3.0 TECHNICAL EVALUATION  


===3.1 Proposed===
3.1     Proposed Change There are several proposed TS changes that either impact NCS analyses or implement changes in fuel storage requirements. The reactor operating condition changes related to the EPU affect fuel depletion, which is credited in the spent fuel pool NCS analysis. EPU-related changes impacting NCS analysis may include higher power density, fuel and moderator temperature changes, and soluble boron concentration changes.
Change There are several proposed TS changes that either impact NCS analyses or implement changes in fuel storage requirements.
One of the TS changes raises the maximum fuel enrichment for fresh fuel storage from a maximum of 4.5 weight-percent uranium-235 to a maximum lattice averaged value of 4.6 weight-percent uranium-235. The TS changes associated with fuel stored in the SFP include increasing the maximum initial enrichment from 4.5 weight-percent uranium-235 to a maximum planar average initial enrichment of 4.6 weight-percent uranium-235, credit for empty storage locations, credit for use of METAMICTM inserts, credit for installation of full-length full-strength, five-fingered control element assemblies and definition of three special configurations referred to in the NCS analysis as inspection and maintenance configurations.
The reactor operating condition changes related to the EPU affect fuel depletion, which is credited in the spent fuel pool NCS analysis.
3.2      Method of Review This safety evaluation involves a review of the NCS analyses for the SFP provided as  to Reference 6 and for the NFV provided as Attachment 4 to Reference 6 and the related proposed TS changes that were provided as attachments to Reference 3. The SFP NCS analysis includes the effects of the changes in reactor operating conditions associated with the EPU and supports expansion of fuel storage requirements to additional storage configurations. The review was performed consistent with Section 9.1.1 of NUREG-0800.
EPU-related changes impacting NCS analysis may include higher power density, fuel and moderator temperature changes, and soluble boron concentration changes. One of the TS changes raises the maximum fuel enrichment for fresh fuel storage from a maximum of 4.5 weight-percent uranium-235 to a maximum lattice averaged value of 4.6 weight-percent uranium-235.
The staff issued an internal memorandum on August 19, 1998, containing guidance for performing the review of SFP NCS analysis (Reference 2). This memorandum is known colloquially as the 'Kopp Memo' (Reference 9), after the author. While the Kopp memorandum does not specify a methodology, it does provide some guidance on the more salient aspects of OFFICIAL USE ONLY      PROPRIETARY INFORM.AJION
The TS changes associated with fuel stored in the SFP include increasing the maximum initial enrichment from 4.5 weight-percent uranium-235 to a maximum planar average initial enrichment of 4.6 weight-percent uranium-235, credit for empty storage locations, credit for use of METAMICTM inserts, credit for installation of full-length full-strength, five-fingered control element assemblies and definition of three special configurations referred to in the NCS analysis as inspection and maintenance configurations.  


===3.2 Method===
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of Review This safety evaluation involves a review of the NCS analyses for the SFP provided as Attachment 3 to Reference 6 and for the NFV provided as Attachment 4 to Reference 6 and the related proposed TS changes that were provided as attachments to Reference
                                                - 4 an NCS analysis, including computer code validation. The guidance is germane to boiling-water reactors and pressurized-water reactors (PWRs), borated and unborated. The Kopp memorandum has been used for virtually every light-water reactor SFP NCS analysis since, including this St. Lucie 2 analysis.
: 3. The SFP NCS analysis includes the effects of the changes in reactor operating conditions associated with the EPU and supports expansion of fuel storage requirements to additional storage configurations.
3.3      SFP NCS Analysis Review 3.3.1   SFP NCS Analysis Method There is no generic or standard methodology for performing NCS analyses for fuel storage and handling. The methods used for the NCS analysis for fuel in the St. Lucie 2 SFP are described in HI-2104753, Revision 4. Additional information describing the methods used is provided in the RAI responses attached to References 4, 5, and 6. Some SFP analysis deficiencies were identified during the review, but as will be discussed below, sufficient margin is built into the analysis methodology to offset the deficiencies. Consequently, the methodology is specific to this analysis and, without further revision, is not appropriate for other applications.
The review was performed consistent with Section 9.1.1 of NUREG-0800.
3.3.1.1 Computational Methods The LAR seeks to credit fuel assembly burnup in 8 of the 10 defined configurations. The CASMO-4 computer code and its 70-group cross-section library were used to calculate burned fuel compositions and to generate lumped-fission-product cross sections for use with the MCNP-5 computer code. MCNP-5 was used, with its ENDF/B-V and ENDF/B-VI continuous energy cross sections and the CASMO-4-generated lumped fission product cross sections, to calculate keff values. These computer codes and the nuclear data sets with them have been used in many NCS analyses, are industry standards, and are considered acceptable.
The staff issued an internal memorandum on August 19, 1998, containing guidance for performing the review of SFP NCS analysis (Reference 2). This memorandum is known colloquially as the 'Kopp Memo' (Reference 9), after the author. While the Kopp memorandum does not specify a methodology, it does provide some guidance on the more salient aspects of OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY
Although not common, the computational method used "lumped fission product" (LFP) number densities and cross sections generated by CASMO for use in MCNP. The preparation, testing, and use of LFP are described in Holtec International report HI-2033031 (Reference 7). The use of LFPs was verified by performing equivalent calculations with CASMO-4 and with MCNP-5 using the LFP number densities and cross sections. These calculations were performed to ensure that CASMO-4 cross sections were appropriately converted to the format required by MCNP-5. There is some unquantified and unvalidated uncertainty associated with using the LFP method. The spent fuel analysis includes a "5 percent of the reactivity decrement" uncertainty (Reference 9) to cover lack of validation of spent fuel compositions, including fission products, and a "[              ] of the minor actinide and fission product worth" uncertainty to cover the lack of validation of minor actinides and fission products for calculation of kef!. Recent work published in NUREG/CR-7109 (Reference 8) indicates that an uncertainty of about 3 percent of the minor actinide and fission product worth should be sufficient to conservatively bound biases that may be associated with calculating keff for systems with minor actinides and fission products. The uncertainties adopted for fuel composition and keff calculations are large enough to cover bias and uncertainty associated with use of LFP.
-an NCS analysis, including computer code validation.
During the review, a question, RAI SRXB-121, was asked concerning how source convergence was checked in the MCNP calculations. According to the RAI response, in all MCNP calculations, results for 50 skipped cycles were used. A review of the Shannon entropy convergence criteria for these calculations showed that in most cases more than 50 skipped cycles would be required to achieve convergence. The RAI response provided an assessment of the impact of revised converged calculations and noted that there was a reduction in the OFFICIAL USE ONLY        PROPRIETARY INFORMATION
The guidance is germane to boiling-water reactors and pressurized-water reactors (PWRs), borated and unborated.
 
The Kopp memorandum has been used for virtually every light-water reactor SFP NCS analysis since, including this St. Lucie 2 analysis.
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3.3 SFP NCS Analysis Review 3.3.1 SFP NCS Analysis Method There is no generic or standard methodology for performing NCS analyses for fuel storage and handling.
                                                  - 5 margin to the keff limits, with the largest reduction being 0.0052 ~k. This nonconservatism is addressed in Section 3.3.5 of this report.
The methods used for the NCS analysis for fuel in the St. Lucie 2 SFP are described in HI-2104753, Revision 4. Additional information describing the methods used is provided in the RAI responses attached to References 4, 5, and 6. Some SFP analysis deficiencies were identified during the review, but as will be discussed below, sufficient margin is built into the analysis methodology to offset the deficiencies.
3.3.1.2 Computer Code Validation Since the NCS analysis credits fuel burnup, it is necessary to consider validation of the computer codes and data used to calculate burned fuel compositions and the computer code and data that utilize the burned fuel compositions to calculate keff for systems with burned fuel.
Consequently, the methodology is specific to this analysis and, without further revision, is not appropriate for other applications.
Consistent with the guidance provided in the Kopp Memo (Reference 9), the analysis has incorporated a "5 percent of the reactivity decrement" uncertainty to cover lack of validation of fuel composition calculations. This uncertainty is calculated as 0.05 times the change in keff from the fresh fuel to the credited final fuel burnup. This uncertainty was calculated by the licensee and applied correctly.
3.3.1.1 Computational Methods The LAR seeks to credit fuel assembly burnup in 8 of the 10 defined configurations.
The study used to support validation of keff calculations using MCNP-5 was documented in Holtec International report HI-2104790 (Reference 10). The validation set included 291 critical configurations that included the French Haut Taux de Combustion (HTC) critical experiments (Reference 11). mixed uranium and plutonium critical experiments and some low enrichment uranium fuel pin experiments. During the review it was noted that the validation set included some of the HTC critical experiments that were not recommended for use, that the trending analysis did not evaluate trends in the calculated bias and bias uncertainty associated with variation of plutonium content or with uranium enrichment, and that the wrong value was used for the fresh fuel bias. Thae licensee performed additional analyses to support its response to RAI SRXB-145, showing the impact of excluding some of the HTC critical experiments, and evaluating the variation of the bias with plutonium content and uranium enrichment. This analysis showed that the bias used in the analysis was about 0.0010 too low. Considering the margins available to the regulatory limit, this 0.001 0 ~k nonconservatism is acceptable.
The CASMO-4 computer code and its 70-group cross-section library were used to calculate burned fuel compositions and to generate lumped-fission-product cross sections for use with the MCNP-5 computer code. MCNP-5 was used, with its ENDF/B-V and ENDF/B-VI continuous energy cross sections and the CASMO-4-generated lumped fission product cross sections, to calculate keff values. These computer codes and the nuclear data sets with them have been used in many NCS analyses, are industry standards, and are considered acceptable.
Appropriate critical experiment data was not available to validate keff calculations crediting minor actinides (uranium-236, neptunium-237, and americium-243) or fission products. This deficiency was identified in the analysis. To address this validation deficiency, an uncertainty equal to [                ] of the worth of the minor actinides and fission products was adopted.
Although not common, the computational method used "lumped fission product" (LFP) number densities and cross sections generated by CASMO for use in MCNP. The preparation, testing, and use of LFP are described in Holtec International report HI-2033031 (Reference 7). The use of LFPs was verified by performing equivalent calculations with CASMO-4 and with MCNP-5 using the LFP number densities and cross sections.
Recent work published in NUREG/CR-7109 indicates that an uncertainty of about 3 percent of the minor actinide and fission product worth should be sufficient to conservatively bound biases that may be associated with minor actinides and fission products. Therefore, the [                ]
These calculations were performed to ensure that CASMO-4 cross sections were appropriately converted to the format required by MCNP-5. There is some unquantified and unvalidated uncertainty associated with using the LFP method. The spent fuel analysis includes a "5 percent of the reactivity decrement" uncertainty (Reference
uncertainty value adequately covers the keff validation deficiencies.
: 9) to cover lack of validation of spent fuel compositions, including fission products, and a "[ ] of the minor actinide and fission product worth" uncertainty to cover the lack of validation of minor actinides and fission products for calculation of kef!. Recent work published in NUREG/CR-7109 (Reference
3.3.2    SFP and Fuel Storage Racks 3.3.2.1 SFP Water Temperature NRC guidance provided in the Kopp memorandum states the NCS analysis should be done at the temperature corresponding to the highest reactivity. Analysis was performed with water densities of 0.9787 g/cm 3 and 1.00 g/cm 3 and at water temperatures of 300 OK and 400 OK and at soluble boron concentrations of 0,500, and 1000 PPM by weight. The water densities used reflect the water density range as the water temperature varies from the temperature where maximum water density occurs, near 39 OF, up to 155 OF, which is the assumed maximum normal operating temperature. The 300 OK and 400 OK temperatures were used because the MCNP continuous energy cross sections are available only at a few specific temperatures and MCNP does not do interpolation or temperature corrections. The product of the analysis was a OFFICIAL USE ONLY          PROPRIETARY INFORMATION
: 8) indicates that an uncertainty of about 3 percent of the minor actinide and fission product worth should be sufficient to conservatively bound biases that may be associated with calculating keff for systems with minor actinides and fission products.
 
The uncertainties adopted for fuel composition and keff calculations are large enough to cover bias and uncertainty associated with use of LFP. During the review, a question, RAI SRXB-121, was asked concerning how source convergence was checked in the MCNP calculations.
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According to the RAI response, in all MCNP calculations, results for 50 skipped cycles were used. A review of the Shannon entropy convergence criteria for these calculations showed that in most cases more than 50 skipped cycles would be required to achieve convergence.
                                                  - 6 storage configuration-dependent temperature bias to be applied to the maximum keff value for that storage configuration.
The RAI response provided an assessment of the impact of revised converged calculations and noted that there was a reduction in the OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION
The NRC staff identified two issues with the SFP water temperature bias determination. The first is that the analysis did not examine water densities between 1.00 and 0.9787 g/cm 3 to ensure that keff did not peak at some value between the maximum and minimum water density values. Instead, engineering judgment was cited by the licensee as the basis for concluding that the calculations were conservative and acceptable. The second issue is that there is currently no way to validate the impact on keff of the variation of cross sections with temperature.
-margin to the keff limits, with the largest reduction being 0.0052 This nonconservatism is addressed in Section 3.3.5 of this report. 3.3.1.2 Computer Code Validation Since the NCS analysis credits fuel burnup, it is necessary to consider validation of the computer codes and data used to calculate burned fuel compositions and the computer code and data that utilize the burned fuel compositions to calculate keff for systems with burned fuel. Consistent with the guidance provided in the Kopp Memo (Reference 9), the analysis has incorporated a "5 percent of the reactivity decrement" uncertainty to cover lack of validation of fuel composition calculations.
There is some uncertainty associated with utilizing 400 OK cross sections to quantify the impact of cross section variation at 342 OK on the keff value calculated by MCNP.
This uncertainty is calculated as 0.05 times the change in keff from the fresh fuel to the credited final fuel burnup. This uncertainty was calculated by the licensee and applied correctly.
From Table 7.3 of HI-21 04753, after discarding bias values that would reduce the maximum keff value, the calculated temperature bias values range up to 0.0034 Ak for case 10 with no soluble boron. It is unlikely that the uncertainty on the temperature corrections is as large as 100 percent. Consequently, for purposes of evaluating a balance of conservatisms and nonconservatisms, the SFP water temperature bias determination may be nonconservative by as much as 0.0034 Ak. This nonconservatism is addressed in Section 3.3.5 of this report.
The study used to support validation of keff calculations using MCNP-5 was documented in Holtec International report HI-2104790 (Reference 10). The validation set included 291 critical configurations that included the French Haut Taux de Combustion (HTC) critical experiments (Reference 11). mixed uranium and plutonium critical experiments and some low enrichment uranium fuel pin experiments.
3.3.2.2 SFP Storage Rack Models Three fuel storage rack variations are used in the St. Lucie 2 SFP. All three are manufactured by fabricating boxes that are then welded together at their corners, creating a checkerboard arrangement of storage cells inside fabricated boxes and between boxes. In response to RAI SRXB-130, the licensee stated that the filler panels and corners that are used to complete the periphery of each rack module have the same thickness as the fabricated boxes and are attached such that the minimum cell inner dimension is maintained.
During the review it was noted that the validation set included some of the HTC critical experiments that were not recommended for use, that the trending analysis did not evaluate trends in the calculated bias and bias uncertainty associated with variation of plutonium content or with uranium enrichment, and that the wrong value was used for the fresh fuel bias. Thae licensee performed additional analyses to support its response to RAI SRXB-145, showing the impact of excluding some of the HTC critical experiments, and evaluating the variation of the bias with plutonium content and uranium enrichment.
The CPR includes BORAL ' panels held to the exterior side of each box using a steel wrapper.
This analysis showed that the bias used in the analysis was about 0.0010 too low. Considering the margins available to the regulatory limit, this 0.001 0 nonconservatism is acceptable.
A detailed representation of the model rack that explicitly models the box, BORAL ' panel, and wrapper was utilized in the analysis. The CPR model used in cases 1 and 8 was a laterally infinite array of fabricated and formed cells. Use of this laterally infinite model, which does not credit neutron leakage from the sides of the CPR rack modules, produces some unquantified margin for the CPR calculations.
Appropriate critical experiment data was not available to validate keff calculations crediting minor actinides (uranium-236, neptunium-237, and americium-243) or fission products.
The Region 1 and 2 rack modules are also constructed steel boxes that are welded together at the corners, creating formed and fabricated storage cells. However, these rack modules do not include BORAL ' panels and wrappers. A simplified model was used for the Region 1 and 2 rack modules that utilizes a single average repeated cell. This model maintains the center-to center spacing between assemblies and the amount of steel between assemblies. This simplification should have little to no effect on the calculated keff values. The only differences between the Region 1 and 2 rack modules is that an L-shaped steel insert has been placed in all Region 1 cells and that L-shaped METAMICTM inserts may be installed in some Region 2 cells. In both the CPR and the Region 2 racks, with METAMICTM inserts, the minimum boron-10 loading is used in the model.
This deficiency was identified in the analysis.
The staff finds that the design basis models used to model the CPR and Region 1 and 2 racks are appropriate and conservative.
To address this validation deficiency, an uncertainty equal to [ ] of the worth of the minor actinides and fission products was adopted. Recent work published in NUREG/CR-7109 indicates that an uncertainty of about 3 percent of the minor actinide and fission product worth should be sufficient to conservatively bound biases that may be associated with minor actinides and fission products.
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Therefore, the [ ] uncertainty value adequately covers the keff validation deficiencies.
 
3.3.2 SFP and Fuel Storage Racks 3.3.2.1 SFP Water Temperature NRC guidance provided in the Kopp memorandum states the NCS analysis should be done at the temperature corresponding to the highest reactivity.
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Analysis was performed with water densities of 0.9787 g/cm 3 and 1.00 g/cm 3 and at water temperatures of 300 OK and 400 OK and at soluble boron concentrations of 0,500, and 1000 PPM by weight. The water densities used reflect the water density range as the water temperature varies from the temperature where maximum water density occurs, near 39 OF, up to 155 OF, which is the assumed maximum normal operating temperature.
                                                  -7 3.3.2.3 SFP Storage Rack Models Manufacturing Tolerances and Uncertainties The minimum material thicknesses of the box walls and sheathing were used in all design basis calculations. Sensitivity calculations were performed for the CPR rack that determined the maximum center-to-center spacing yielded higher keff values. For the Region 1 and 2 rack modules, the minimum cell inner dimension was used, which is conservative and therefore, acceptable.
The 300 OK and 400 OK temperatures were used because the MCNP continuous energy cross sections are available only at a few specific temperatures and MCNP does not do interpolation or temperature corrections.
3.3.3    Fuel Assembly 3.3.3.1 Bounding Fuel Assembly Design The fuel assemblies used at St. Lucie 2 are all Combustion Engineering (CE) 16x16 or equivalent assemblies manufactured by other suppliers. In some of the older CE 16x16 assemblies, fuel rods were replaced with rods containing B4 C. More recently, the fuel assemblies have included some fuel rods with Gd 2 0 3 mixed in with the U02 in the fuel pellets.
The product of the analysis was a OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY
When Gd 20 3 is mixed in the fuel pellets with the U02 , the uranium-235 enrichment is lower than the maximum planar average enrichment for the fuel pellets that do not have Gd 2 0 3 . In both new and old designs, the initial uranium enrichment may be varied from pin-to-pin to control fuel assembly power peaking factors. The fuel assembly design model used in the St. Lucie 2 SFP NCS analysis does not include any B4 C or Gd 2 0 3 rods and utilizes the maximum planar average enrichment for all fuel pins. This increases the amount of uranium-235 present relative to the actual fuel assemblies. Sensitivity calculations were performed to estimate the bias introduced by using the planar average enrichment rather than the actual pin-dependent enrichment variations. The licensee performed sensitivity studies that indicate that modeling the fuel assemblies as having all fuel pellets at the maximum planar average enrichment rather than explicitly modeling the Gd 2 0 3 fuel rods and the replacement of some fuel rods with B4 C rods adds additional unquantified margin to the analysis.
-storage configuration-dependent temperature bias to be applied to the maximum keff value for that storage configuration.
Some fuel assemblies stored at St. Lucie 2 include 6-inch 2.6 weight-percent enriched uranium-235 blankets on the ends of the fuel assemblies. The licensee's calculations were performed both with and without enriched uranium blankets to ensure that the most reactive design was used for each case and, where appropriate, each point on each burnup dependent loading curve. Blankets of different sizes and enrichment were not included in the analysis and therefore are not part of the methodology.
The NRC staff identified two issues with the SFP water temperature bias determination.
The fuel assembly model included only the active length of the fuel rods. The bounding model did not include fuel assembly grids, nozzles, or the nonfuel ends of the fuel rods. The licensee's sensitivity calculations were performed both with and without soluble boron to show that modeling the grids either reduced reactivity or, when soluble boron was present, caused only a small increase in keff. Not modeling the grids is conservative at unborated conditions, but at some point the amount of soluble boron in the water would make it nonconservative to not model the grids. At 500 ppm of soluble boron, the value used for normal conditions with soluble boron, the impact may be as large as 0.0008 ~k. It is expected that this value will increase as soluble boron concentrations are increased for accident conditions. However, this increase in
The first is that the analysis did not examine water densities between 1.00 and 0.9787 g/cm 3 to ensure that keff did not peak at some value between the maximum and minimum water density values. Instead, engineering judgment was cited by the licensee as the basis for concluding that the calculations were conservative and acceptable.
~k is offset by the residual uncredited soluble boron.
The second issue is that there is currently no way to validate the impact on keff of the variation of cross sections with temperature.
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There is some uncertainty associated with utilizing 400 OK cross sections to quantify the impact of cross section variation at 342 OK on the keff value calculated by MCNP. From Table 7.3 of HI-21 04753, after discarding bias values that would reduce the maximum keff value, the calculated temperature bias values range up to 0.0034 Ak for case 10 with no soluble boron. It is unlikely that the uncertainty on the temperature corrections is as large as 100 percent. Consequently, for purposes of evaluating a balance of conservatisms and nonconservatisms, the SFP water temperature bias determination may be nonconservative by as much as 0.0034 Ak. This nonconservatism is addressed in Section 3.3.5 of this report. 3.3.2.2 SFP Storage Rack Models Three fuel storage rack variations are used in the St. Lucie 2 SFP. All three are manufactured by fabricating boxes that are then welded together at their corners, creating a checkerboard arrangement of storage cells inside fabricated boxes and between boxes. In response to RAI SRXB-130, the licensee stated that the filler panels and corners that are used to complete the periphery of each rack module have the same thickness as the fabricated boxes and are attached such that the minimum cell inner dimension is maintained.
 
The CPR includes BORAL Žpanels held to the exterior side of each box using a steel wrapper. A detailed representation of the model rack that explicitly models the box, BORAL Ž panel, and wrapper was utilized in the analysis.
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The CPR model used in cases 1 and 8 was a laterally infinite array of fabricated and formed cells. Use of this laterally infinite model, which does not credit neutron leakage from the sides of the CPR rack modules, produces some unquantified margin for the CPR calculations.
                                                  -8 3.3.3.2 Fuel Assembly Manufacturing Tolerances and Uncertainties The licensee used an uncommon technique to address fuel storage rack and fuel assembly manufacturing tolerances. In its approach, the licensee used the worst case dimensions for key parameters in the model, effectively incorporating the uncertainty as a bias. It then performed an analysis to show that the margin provided by using the bounding values for a few key parameters was larger than the margin produced by doing detailed uncertainty analysis around the nominal conditions, combining the uncertainty contributions. According to the analysis presented in HI-2104753, including worst case dimensions produces conservative margins varying from 0.0014 to 0.0163 ~k compared to using nominal dimensions and including a more conventional detailed uncertainty analysis. Due to the basis for the method and range of margins observed for the St. Lucie 2 analysis, any future licensing basis changes utilizing this method will require the same comparison of detailed uncertainty analysis results to the simplified method results. For the St. Lucie 2 analysis, the method does generate reasonably conservative margins compared to the more conventional detailed uncertainty analysis.
The Region 1 and 2 rack modules are also constructed steel boxes that are welded together at the corners, creating formed and fabricated storage cells. However, these rack modules do not include BORAL Ž panels and wrappers.
3.3.3.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on uranium-235 enrichment and various manufacturing tolerances. The manufacturing tolerances are typically manifested as uncertainties, as discussed above, or are bounded by values used in the analysis. These tolerances and bounding values would also carry through to the spent nuclear fuel, common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. The characterization of spent nuclear fuel is more problematic. Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three main areas: a burnup uncertainty, the axial and radial apportionment of the burnup, and the core operation that achieved that burnup.
A simplified model was used for the Region 1 and 2 rack modules that utilizes a single average repeated cell. This model maintains the center spacing between assemblies and the amount of steel between assemblies.
3.3.3.4 Burnup Uncertainty In the Kopp Memo (Reference 9), the NRC staff provided its method for evaluating burnup uncertainty, "A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculational uncertainties. In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption." The licensee used this approach in HI-2104753 to address the uncertainty in the burned fuel compositions.
This simplification should have little to no effect on the calculated keff values. The only differences between the Region 1 and 2 rack modules is that an L-shaped steel insert has been placed in all Region 1 cells and that L-shaped METAMICTM inserts may be installed in some Region 2 cells. In both the CPR and the Region 2 racks, with METAMICTM inserts, the minimum boron-10 loading is used in the model. The staff finds that the design basis models used to model the CPR and Region 1 and 2 racks are appropriate and conservative.
3.3.3.5 Axial Apportionment of the Burnup or Axial Burnup Profile Another important aspect of fuel characterization is the selection of the axial burnup profile. At the beginning of life, a PWR fuel assembly will be exposed to a near-cosine axial-shaped flux, which will deplete fuel near the axial center at a greater rate than at the ends. As the reactor continues to operate, the cosine flux shape will flatten because of the fuel depletion and fission-product buildup that occurs near the center. Near the fuel assembly ends, burnup is suppressed due to leakage. If a uniform axial burnup profile is assumed, then the burnup at the ends is over predicted. Analysis discussed in NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis" (Reference 12), has shown that, at assembly burnups above about 10 to 20 GWd/MTU, this results in an underprediction of keff; generally the underprediction becomes larger as burnup increases. This is what is known as the "end effect." Proper selection of the axial burnup profile is necessary to ensure keff is not underpredicted due to the end effect.
OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION -7 3.3.2.3 SFP Storage Rack Models Manufacturing Tolerances and Uncertainties The minimum material thicknesses of the box walls and sheathing were used in all design basis calculations.
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Sensitivity calculations were performed for the CPR rack that determined the maximum center-to-center spacing yielded higher keff values. For the Region 1 and 2 rack modules, the minimum cell inner dimension was used, which is conservative and therefore, acceptable.
 
3.3.3 Fuel Assembly 3.3.3.1 Bounding Fuel Assembly Design The fuel assemblies used at St. Lucie 2 are all Combustion Engineering (CE) 16x16 or equivalent assemblies manufactured by other suppliers.
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In some of the older CE 16x16 assemblies, fuel rods were replaced with rods containing B 4 C. More recently, the fuel assemblies have included some fuel rods with Gd 2 0 3 mixed in with the U0 2 in the fuel pellets. When Gd 2 0 3 is mixed in the fuel pellets with the U0 2 , the uranium-235 enrichment is lower than the maximum planar average enrichment for the fuel pellets that do not have Gd 2 0 3. In both new and old designs, the initial uranium enrichment may be varied from pin-to-pin to control fuel assembly power peaking factors. The fuel assembly design model used in the St. Lucie 2 SFP NCS analysis does not include any B 4 C or Gd 2 0 3 rods and utilizes the maximum planar average enrichment for all fuel pins. This increases the amount of uranium-235 present relative to the actual fuel assemblies.
                                                  - 9 NUREG/CR-6801 provides recommendations for selecting an appropriate axial burnup profile.
Sensitivity calculations were performed to estimate the bias introduced by using the planar average enrichment rather than the actual pin-dependent enrichment variations.
With respect to the axial burnup profile, HI-21 04753 did not use the axial burnup profiles from NUREG/CR-6801. A description of how the axial burnup profiles were derived is provided in Sections 2.3.5.6 and 7.11 of HI-2104753. The NRC staff requested additional information regarding the derivation of the axial burnup profiles, which the licensee provided in its response to RAI SRXB-139 (Reference 6).
The licensee performed sensitivity studies that indicate that modeling the fuel assemblies as having all fuel pellets at the maximum planar average enrichment rather than explicitly modeling the Gd 2 0 3 fuel rods and the replacement of some fuel rods with B 4 C rods adds additional unquantified margin to the analysis.
The axial burnup profiles used were derived from 744 St. Lucie 2 fuel assembly burnup profiles.
Some fuel assemblies stored at St. Lucie 2 include 6-inch 2.6 weight-percent enriched uranium-235 blankets on the ends of the fuel assemblies.
These profiles include 620 pre-EPU profiles and 124 profiles from design calculations for post-EPU cycles. The 744 profiles include 558 profiles for assemblies with enriched uranium blankets and 188 profiles for non blanketed fuel. Separate profiles were prepared for blanketed and unblanketed fuel. Blanketed and unblanketed fuel assemblies are treated differently because application of the relatively lower burnups associated with axial blanket zones to unblanketed assemblies would be significantly and unnecessarily over-conservative.
The licensee's calculations were performed both with and without enriched uranium blankets to ensure that the most reactive design was used for each case and, where appropriate, each point on each burnup dependent loading curve. Blankets of different sizes and enrichment were not included in the analysis and therefore are not part of the methodology.
The fuel assembly model included only the active length of the fuel rods. The bounding model did not include fuel assembly grids, nozzles, or the nonfuel ends of the fuel rods. The licensee's sensitivity calculations were performed both with and without soluble boron to show that modeling the grids either reduced reactivity or, when soluble boron was present, caused only a small increase in keff. Not modeling the grids is conservative at unborated conditions, but at some point the amount of soluble boron in the water would make it nonconservative to not model the grids. At 500 ppm of soluble boron, the value used for normal conditions with soluble boron, the impact may be as large as 0.0008 It is expected that this value will increase as soluble boron concentrations are increased for accident conditions.
However, this increase in is offset by the residual uncredited soluble boron. OFFICI,'\L USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION -8 3.3.3.2 Fuel Assembly Manufacturing Tolerances and Uncertainties The licensee used an uncommon technique to address fuel storage rack and fuel assembly manufacturing tolerances.
In its approach, the licensee used the worst case dimensions for key parameters in the model, effectively incorporating the uncertainty as a bias. It then performed an analysis to show that the margin provided by using the bounding values for a few key parameters was larger than the margin produced by doing detailed uncertainty analysis around the nominal conditions, combining the uncertainty contributions.
According to the analysis presented in HI-2104753, including worst case dimensions produces conservative margins varying from 0.0014 to 0.0163 compared to using nominal dimensions and including a more conventional detailed uncertainty analysis.
Due to the basis for the method and range of margins observed for the St. Lucie 2 analysis, any future licensing basis changes utilizing this method will require the same comparison of detailed uncertainty analysis results to the simplified method results. For the St. Lucie 2 analysis, the method does generate reasonably conservative margins compared to the more conventional detailed uncertainty analysis.
3.3.3.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on uranium-235 enrichment and various manufacturing tolerances.
The manufacturing tolerances are typically manifested as uncertainties, as discussed above, or are bounded by values used in the analysis.
These tolerances and bounding values would also carry through to the spent nuclear fuel, common industry practice has been to treat the uncertainties as unaffected by the fuel depletion.
The characterization of spent nuclear fuel is more problematic.
Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three main areas: a burnup uncertainty, the axial and radial apportionment of the burnup, and the core operation that achieved that burnup. 3.3.3.4 Burnup Uncertainty In the Kopp Memo (Reference 9), the NRC staff provided its method for evaluating burnup uncertainty, "A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculational uncertainties.
In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption." The licensee used this approach in HI-2104753 to address the uncertainty in the burned fuel compositions.
3.3.3.5 Axial Apportionment of the Burnup or Axial Burnup Profile Another important aspect of fuel characterization is the selection of the axial burnup profile. At the beginning of life, a PWR fuel assembly will be exposed to a near-cosine axial-shaped flux, which will deplete fuel near the axial center at a greater rate than at the ends. As the reactor continues to operate, the cosine flux shape will flatten because of the fuel depletion and fission-product buildup that occurs near the center. Near the fuel assembly ends, burnup is suppressed due to leakage. If a uniform axial burnup profile is assumed, then the burnup at the ends is over predicted.
Analysis discussed in NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis" (Reference 12), has shown that, at assembly burnups above about 10 to 20 GWd/MTU, this results in an underprediction of k eff; generally the underprediction becomes larger as burnup increases.
This is what is known as the "end effect." Proper selection of the axial burnup profile is necessary to ensure keff is not underpredicted due to the end effect. OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY
-NUREG/CR-6801 provides recommendations for selecting an appropriate axial burnup profile. With respect to the axial burnup profile, HI-21 04753 did not use the axial burnup profiles from NUREG/CR-6801.
A description of how the axial burnup profiles were derived is provided in Sections 2.3.5.6 and 7.11 of HI-2104753.
The NRC staff requested additional information regarding the derivation of the axial burnup profiles, which the licensee provided in its response to RAI SRXB-139 (Reference 6). The axial burnup profiles used were derived from 744 St. Lucie 2 fuel assembly burnup profiles.
These profiles include 620 pre-EPU profiles and 124 profiles from design calculations for post-EPU cycles. The 744 profiles include 558 profiles for assemblies with enriched uranium blankets and 188 profiles for non blanketed fuel. Separate profiles were prepared for blanketed and un blanketed fuel. Blanketed and unblanketed fuel assemblies are treated differently because application of the relatively lower burnups associated with axial blanket zones to unblanketed assemblies would be significantly and unnecessarily over-conservative.
The process described in HI-21 04753 and the RAI SRXB-139 response results in relative burnup-dependent axial burnup profiles that are more reactive than all of the 744 profiles. One conservative feature of the bounding axial burnup profiles is that they are not renormalized to yield an assembly average relative burnup of 1.0. Instead, use of the profiles conservatively reduces the assembly average burnup by 1.2 to 3.5 percent. The use of these conservative fuel-assembly-design-specific and plant-specific axial burnup profiles is acceptable.
The process described in HI-21 04753 and the RAI SRXB-139 response results in relative burnup-dependent axial burnup profiles that are more reactive than all of the 744 profiles. One conservative feature of the bounding axial burnup profiles is that they are not renormalized to yield an assembly average relative burnup of 1.0. Instead, use of the profiles conservatively reduces the assembly average burnup by 1.2 to 3.5 percent. The use of these conservative fuel-assembly-design-specific and plant-specific axial burnup profiles is acceptable.
The burnup credit limit curves were derived using polynomial fits that bound the most limiting result obtained using a uniform profile, nonblanket profile and a blanketed fuel profile. Consequently, separate loading curves for blanketed and nonblanketed fuel are not needed. 3.3.3.6 Planar Burnup Distribution Due to the neutron flux gradients in the reactor core, assemblies can show a radially tilted burnup distribution (Le., differences in burnup between portions or quadrants of the cross section of the assembly).
The burnup credit limit curves were derived using polynomial fits that bound the most limiting result obtained using a uniform profile, nonblanket profile and a blanketed fuel profile.
The HI-21 04753 analysis did not consider the effect of planar burnup distribution on reactivity.
Consequently, separate loading curves for blanketed and nonblanketed fuel are not needed.
The impact of radial burnup gradients may be estimated by comparing the distribution of radial burnup tilt information provided in Figure 3-4 of DOE/RW-0496 (Reference
3.3.3.6 Planar Burnup Distribution Due to the neutron flux gradients in the reactor core, assemblies can show a radially tilted burnup distribution (Le., differences in burnup between portions or quadrants of the cross section of the assembly). The HI-21 04753 analysis did not consider the effect of planar burnup distribution on reactivity. The impact of radial burnup gradients may be estimated by comparing the distribution of radial burnup tilt information provided in Figure 3-4 of DOE/RW-0496 (Reference 13) with information on the sensitivity of keff to radial burnup tilt provided in Section 6.1.2 of NllREG/CR-6800 (Reference 14). From DOE/RW-0496, the maximum quadrant deviation from assembly average burnup had been observed to be less than 25 percent at low (burnup < 20 GWdlMTU) assembly average burnups and was observed to decrease with burnup, generally being less than 10 percent at burnups above 20 GWd/MTU.
: 13) with information on the sensitivity of keff to radial burnup tilt provided in Section 6.1.2 of NllREG/CR-6800 (Reference 14). From DOE/RW-0496, the maximum quadrant deviation from assembly average burnup had been observed to be less than 25 percent at low (burnup < 20 GWdlMTU) assembly average burnups and was observed to decrease with burnup, generally being less than 10 percent at burnups above 20 GWd/MTU. Combining these radial tilt bounding estimates with the keff sensitivity information provided in NllREG/CR-6800, the NRC staff's review of the radial burnup tilts could raise the keff value as much as 0.002 ilk. With the information available, the staff finds that it is conservative to consider this effect as a bias, its potential impact is small, and it is accommodated within the analysis margins. 3.3.3.7 Burnup History/Core Operating Parameters NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," (Reference
Combining these radial tilt bounding estimates with the keff sensitivity information provided in NllREG/CR-6800, the NRC staff's review of the radial burnup tilts could raise the keff value as much as 0.002 ilk. With the information available, the staff finds that it is conservative to consider this effect as a bias, its potential impact is small, and it is accommodated within the analysis margins.
: 15) provides some discussion on the treatment of depletion analysis parameters that determine how the burnup was achieved.
3.3.3.7 Burnup History/Core Operating Parameters NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," (Reference 15) provides some discussion on the treatment of depletion analysis parameters that determine how the burnup was achieved. While NUREG/CR-6665 is focused on NCS analysis in storage and transportation casks, the basic prinCipals with respect to the depletion analysis apply generically, since the phenomena occur in the reactor as the fuel is OFFICIAL USE ONLY         PROPRIETi\RY INFORMATION
While NUREG/CR-6665 is focused on NCS analysis in storage and transportation casks, the basic prinCipals with respect to the depletion analysis apply generically, since the phenomena occur in the reactor as the fuel is OFFICIAL USE ONLY PROPRIETi\RY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION  
 
-being used. The results have some applicability to St. Lucie 2 NCS analyses.
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The basic strategy is to select parameters that maximize the Doppler broadening/spectral hardening of the neutron field resulting in maximum plutonium-239/241 production.
                                                - 10 being used. The results have some applicability to St. Lucie 2 NCS analyses. The basic strategy is to select parameters that maximize the Doppler broadening/spectral hardening of the neutron field resulting in maximum plutonium-239/241 production. NUREG/CR-6665 discusses six parameters affecting the depletion analysis: fuel temperature, moderator temperature, soluble boron, specific power and operating history, fixed burnable poisons, and integral burnable poisons. While the mechanism for each is different, the effect is similar: Doppler broadening/spectral hardening of the neutron field resulting in increased plutonium-239/241 production. NUREG/CR-6665 provides an estimate of the reactivity worth of these parameters.
NUREG/CR-6665 discusses six parameters affecting the depletion analysis:
The largest effect appears to be due to moderator temperature. NUREG/CR-6665 approximates the moderator temperature effect, in an infinite lattice of high burnup fuel, to be 90 pcmfOK. Thus, a 10&deg;F change in moderator temperature used in the depletion analysis would result in 0.005 ~keff' The effects of each core operating parameter typically have a burnup or time dependency.
fuel temperature, moderator temperature, soluble boron, specific power and operating history, fixed burnable poisons, and integral burnable poisons. While the mechanism for each is different, the effect is similar: Doppler broadening/spectral hardening of the neutron field resulting in increased plutonium-239/241 production.
For fuel and moderator temperatures, NUREG/CR-6665 recommends using the maximum operating temperatures to maximize plutonium-239/241 production. For moderator temperature, the HI-21 04753 analysis used the post-EPU exit water temperature for the peak power assembly and used a conservatively high fuel temperature. The moderator and fuel temperatures used are therefore acceptable.
NUREG/CR-6665 provides an estimate of the reactivity worth of these parameters.
For boron concentration, NUREG/CR-6665 recommends using a conservatively high cycle-average boron concentration. The licensee's analysis used a cycle average soluble boron concentration of 750 ppm for pre-EPU cycles and 1000 ppm for post-EPU cycles. The data for S1. Lucie 2 cycles 13 through 18 were used to correctly calculate the average soluble boron concentration for each cycle, yielding a maximum pre-EPU value of 650 ppm, which is well below the 750 ppm value used for pre-EPU fuel burnup calculations. For post-EPU cycles, a value of 1000 ppm was adopted. This value is 300 ppm higher than the predicted boron letdown curve provided in Table F.8 of HI-21 04753 and is to bound future operations. The boron concentrations used are therefore acceptable.
The largest effect appears to be due to moderator temperature.
Specific power and operating history are related. Operating history is essentially the history of time spent at various specific power levels. Specific power is a second order effect compared to moderator temperature and soluble boron concentration. The HI-21 04753 analysis used a pre-EPU specific power of 33.5 MW/MTU and a post-EPU specific power of 37.5 MW/MTU. As is described in Section 7.4 of HI-21 04753 and presented in Tables 7.25 through 7.31, sensitivity calculations were performed for +/- 5 MW/MTU. These calculations demonstrated that the St. Lucie 2 SFP rack keff values are only weakly dependent on the speCific power level. The uncertainties associated with specific power and operating history are very small compared to the other uncertainties explicitly included in the analysis and therefore have negligible impact on the overall uncertainty. The specific power and operating history used are therefore acceptable.
NUREG/CR-6665 approximates the moderator temperature effect, in an infinite lattice of high burnup fuel, to be 90 pcmfOK. Thus, a 10&deg;F change in moderator temperature used in the depletion analysis would result in 0.005 The effects of each core operating parameter typically have a burnup or time dependency.
3.3.3.8 Integral and Fixed Burnable Absorbers St. Lucie 2 has in the past used fuel assemblies in which some of the fuel rods were replaced with rods filled with B4 C. St. Lucie 2 has also used fuel rods in which some of the rods contained fuel plus Gd 20 3
For fuel and moderator temperatures, NUREG/CR-6665 recommends using the maximum operating temperatures to maximize plutonium-239/241 production.
* Rather than explicitly modeling the B4C or U02+Gd 20 3 fuel rods, the HI-2104753 analysis used models where all fuel rods were assumed to be at maximum planar average enrichment. As discussed in Section 3.3.3.1 of this evaluation sensitivity studies for St. Lucie 2 SFP indicate this is a conservative modeling simplification.
For moderator temperature, the HI-21 04753 analysis used the post-EPU exit water temperature for the peak power assembly and used a conservatively high fuel temperature.
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The moderator and fuel temperatures used are therefore acceptable.
 
For boron concentration, NUREG/CR-6665 recommends using a conservatively high cycle-average boron concentration.
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The licensee's analysis used a cycle average soluble boron concentration of 750 ppm for pre-EPU cycles and 1000 ppm for post-EPU cycles. The data for S1. Lucie 2 cycles 13 through 18 were used to correctly calculate the average soluble boron concentration for each cycle, yielding a maximum pre-EPU value of 650 ppm, which is well below the 750 ppm value used for pre-EPU fuel burnup calculations.
                                                - 11 Removable neutron absorbing rods have not been used in the St. Lucie 2 reactor.
For post-EPU cycles, a value of 1000 ppm was adopted. This value is 300 ppm higher than the predicted boron letdown curve provided in Table F.8 of HI-21 04753 and is to bound future operations.
3.3.4   Determination of Soluble Boron Requirements Section 50.68 of 10 CFR requires that the keff of the St. Lucie 2 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 unborated water. By considering the double contingency principal, the 1900 ppm of soluble boron that is required by the St. Lucie 2 post-EPU TS can be credited to ensure the St. Lucie 2 keff does not exceed 0.95, provided that the accident and a boron dilution are independent events. Since the minimum required soluble boron has increased from 1720 ppm to 1900 ppm and the soluble boron credited under normal conditions has decreased from 520 to 500 ppm, the pre-EPU soluble boron dilution analysis bounds the post-EPU configuration.
The boron concentrations used are therefore acceptable.
HI-21 04753 considered the following accidents: abnormal temperature, dropped, mislocated, and misloaded fuel assemblies, missing or damaged required METAMICTM insert, missing required control rods, use of an incorrect loading curve, misalignment between the active fuel region and the neutron absorber. The licensee's analysis demonstrated that the fuel storage racks will be subcritical under accident conditions with 1500 ppm of soluble boron, which is below the revised TS required 1900 ppm of soluble boron.
Specific power and operating history are related. Operating history is essentially the history of time spent at various specific power levels. Specific power is a second order effect compared to moderator temperature and soluble boron concentration.
3.3.5   Margin Analysis and Comparison with Remaining Uncertainties This section provides evaluation of additional conservatism in the analysis and evaluation of items that may have been treated nonconservatively.
The HI-21 04753 analysis used a pre-EPU specific power of 33.5 MW/MTU and a post-EPU specific power of 37.5 MW/MTU. As is described in Section 7.4 of HI-21 04753 and presented in Tables 7.25 through 7.31, sensitivity calculations were performed for +/- 5 MW/MTU. These calculations demonstrated that the St. Lucie 2 SFP rack keff values are only weakly dependent on the speCific power level. The uncertainties associated with specific power and operating history are very small compared to the other uncertainties explicitly included in the analysis and therefore have negligible impact on the overall uncertainty.
3.3.5.1 Potential Nonconservatisms The response to RAI SRXB-121 (Reference 6) indicates vthat results from some nonconverged MCNP calculations were used in the analysis. Those new calculations improved convergence and the case-specific uncertainty increased by 0.001 ~k for some cases. This small increase is covered by the conservative margins described below. The nonconverged cases also affected the determination of the loading curves, increasing the keff by up to 0.0052 ~k for some of the points. Since the curves were generated using a target keff + bias + uncertainties of 0.99, the keff of the normal conditions associated with the loading curves may be as high as 0.9952.
The specific power and operating history used are therefore acceptable.
Considering the conservative margins described below, the NRC staff finds this acceptable.
3.3.3.8 Integral and Fixed Burnable Absorbers St. Lucie 2 has in the past used fuel assemblies in which some of the fuel rods were replaced with rods filled with B 4 C. St. Lucie 2 has also used fuel rods in which some of the rods contained fuel plus Gd 2 0 3* Rather than explicitly modeling the B 4 C or U0 2+Gd 2 0 3 fuel rods, the HI-2104753 analysis used models where all fuel rods were assumed to be at maximum planar average enrichment.
As was noted in Section 3.3.3.6, the evaluation of the planar burnup distribution could increase the estimated keff from the St. Lucie 2 NCS analysis by as much as 0.002 ~k.
As discussed in Section 3.3.3.1 of this evaluation sensitivity studies for St. Lucie 2 SFP indicate this is a conservative modeling simplification.
As was noted in Section 3.3.2.1, the evaluation of the temperature dependence of the calculated keff values did not look at water densities and temperature between the normal condition extremes. It is possible that the calculated keff values may, in some cases, have peaked between the extremes. It is unlikely that the nonconservatism, if it exists, could be as large as 0.0034 ~k. Considering the conservative margins described below, the NRC staff finds this acceptable.
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The analysis incorporates a 2.5-percent uncertainty in the fuel assembly average burnup. The response to RAI SRXB-128 (Reference 5) indicates that 1.5 percent of the 2.5 percent covers uncertainty in plant secondary calorimetric power. This leaves 2-percent uncertainty on the assembly relative power. Uncertainty is introduced because only approximately 25 percent of OFFICIAL USE ONLY       PROPRIETARY INFORMATION
-11 Removable neutron absorbing rods have not been used in the St. Lucie 2 reactor. 3.3.4 Determination of Soluble Boron Requirements Section 50.68 of 10 CFR requires that the keff of the St. Lucie 2 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 unborated water. By considering the double contingency principal, the 1900 ppm of soluble boron that is required by the St. Lucie 2 post-EPU TS can be credited to ensure the St. Lucie 2 keff does not exceed 0.95, provided that the accident and a boron dilution are independent events. Since the minimum required soluble boron has increased from 1720 ppm to 1900 ppm and the soluble boron credited under normal conditions has decreased from 520 to 500 ppm, the pre-EPU soluble boron dilution analysis bounds the post-EPU configuration.
 
HI-21 04753 considered the following accidents:
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abnormal temperature, dropped, mislocated, and misloaded fuel assemblies, missing or damaged required METAMICTM insert, missing required control rods, use of an incorrect loading curve, misalignment between the active fuel region and the neutron absorber.
                                                  - 12 the assemblies are instrumented and that there are uncertainties associated with neutron flux measurement, detector calibration, integration of power over the life of the assembly, and extrapolation of measured data to uninstrumented assemblies. A value more typically used is 5 percent of burnup. Because there is conservatism in the derivation of the axial burnup profiles, and the other conservatisms described in the next subsection, use of the low 2.5-percent uncertainty in the St. Lucie 2 NCS analysis is, therefore, acceptable in the context of the available margins.
The licensee's analysis demonstrated that the fuel storage racks will be subcritical under accident conditions with 1500 ppm of soluble boron, which is below the revised TS required 1900 ppm of soluble boron. 3.3.5 Margin Analysis and Comparison with Remaining Uncertainties This section provides evaluation of additional conservatism in the analysis and evaluation of items that may have been treated nonconservatively.
3.3.5.2 Potential Analysis Conservatisms The analysis includes several aspects that add margin to the analysis.
3.3.5.1 Potential Nonconservatisms The response to RAI SRXB-121 (Reference
These include the following:
: 6) indicates vthat results from some nonconverged MCNP calculations were used in the analysis.
* Axial burnup profiles not renormalized Application of the axial burnup profiles artificially reduces the assembly average burnups by 1.2 to 3.5 percent. This represents additional margin of about 0.003 L\k for an unblanketed assembly burned to 50 GWd/MTU.
Those new calculations improved convergence and the case-specific uncertainty increased by 0.001 for some cases. This small increase is covered by the conservative margins described below. The nonconverged cases also affected the determination of the loading curves, increasing the keff by up to 0.0052 for some of the points. Since the curves were generated using a target keff + bias + uncertainties of 0.99, the keff of the normal conditions associated with the loading curves may be as high as 0.9952. Considering the conservative margins described below, the NRC staff finds this acceptable.
* Burnup coefficient loading curve defined at keff + uncertainties + biases equal to 0.99 Section 50.68 of 10 CFR requires that, when soluble boron in the spent fuel pool is credited, the keff must be below 1.0 when no soluble boron is present. Generating loading curves with a keff of 0.99, without soluble boron, includes 0.01 L\k margin to the analysis.
As was noted in Section 3.3.3.6, the evaluation of the planar burnup distribution could increase the estimated keff from the St. Lucie 2 NCS analysis by as much as 0.002 As was noted in Section 3.3.2.1, the evaluation of the temperature dependence of the calculated keff values did not look at water densities and temperature between the normal condition extremes.
* Burnup coefficient loading curve fits either match or conservatively bound the calculated burnup coefficient curve points The final loading curves are second-order polynomial fits that conservatively bound the calculated acceptable loading points. This introduces a small amount of additional margin for most of the range.
It is possible that the calculated keff values may, in some cases, have peaked between the extremes.
* Conservative treatment of uncertainties An unusual treatment of manufacturing tolerances was used that, according to the analysis, introduced additional margin varying from 0.0014 to 0.0163 L\k.
It is unlikely that the nonconservatism, if it exists, could be as large as 0.0034 Considering the conservative margins described below, the NRC staff finds this acceptable.
* Modeling all fuel rods at the maximum planar average enrichment rather than explicitly modeling B4 C or U02 +Gd2 0 3 fuel rods Many of the fuel assemblies include either fuel rods with U02+Gd 20 3 pellets or with B4 C rods replacing fuel rods. The presence of these Gd rods and B4 C rods is not credited in the analysis. This represents significant unquantified margin. From a generic study presented in Reference 15, not modeling the Gd20 3 represents margin ranging from about 0.08 L\k at zero burnup to 0.002 L\k at high burnups. From Figure 36 in the Reference 15, not modeling the B4 C rods in a CE 14x14 assembly results in margin ranging from around 0.15 L\k at low burnups to 0.002 L\k at burnups around 30 GWdlMTU.
The analysis incorporates a 2.5-percent uncertainty in the fuel assembly average burnup. The response to RAI SRXB-128 (Reference
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: 5) indicates that 1.5 percent of the 2.5 percent covers uncertainty in plant secondary calorimetric power. This leaves 2-percent uncertainty on the assembly relative power. Uncertainty is introduced because only approximately 25 percent of OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY  
 
-the assemblies are instrumented and that there are uncertainties associated with neutron flux measurement, detector calibration, integration of power over the life of the assembly, and extrapolation of measured data to uninstrumented assemblies.
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A value more typically used is 5 percent of burnup. Because there is conservatism in the derivation of the axial burnup profiles, and the other conservatisms described in the next subsection, use of the low 2.5-percent uncertainty in the St. Lucie 2 NCS analysis is, therefore, acceptable in the context of the available margins. 3.3.5.2 Potential Analysis Conservatisms The analysis includes several aspects that add margin to the analysis.
                                                -13
These include the following: Axial burnup profiles not renormalized Application of the axial burnup profiles artificially reduces the assembly average burnups by 1.2 to 3.5 percent. This represents additional margin of about 0.003 L\k for an unblanketed assembly burned to 50 GWd/MTU.
* An uncertainty equal to [               ] of the minor actinide and fission product worth was used Recent work documented in Reference 8 indicates that use of an uncertainty as low as 1.S percent of the minor actinide and fission product worth may be appropriate. Use of the [               ] uncertainty adds margin to the uncertainty analysis. The
* Burnup coefficient loading curve defined at keff + uncertainties  
        "[             ] of the worth" uncertainty results in uncertainties ranging from 0.00S4 ilk at low burnups to 0.026 ilk at high burnups. At high burnups, use of the [               ]
+ biases equal to 0.99 Section 50.68 of 10 CFR requires that, when soluble boron in the spent fuel pool is credited, the keff must be below 1.0 when no soluble boron is present. Generating loading curves with a keff of 0.99, without soluble boron, includes 0.01 L\k margin to the analysis. Burnup coefficient loading curve fits either match or conservatively bound the calculated burnup coefficient curve points The final loading curves are second-order polynomial fits that conservatively bound the calculated acceptable loading points. This introduces a small amount of additional margin for most of the range. Conservative treatment of uncertainties An unusual treatment of manufacturing tolerances was used that, according to the analysis, introduced additional margin varying from 0.0014 to 0.0163 L\k. Modeling all fuel rods at the maximum planar average enrichment rather than explicitly modeling B 4 C or U0 2+Gd 2 0 3 fuel rods Many of the fuel assemblies include either fuel rods with U0 2+Gd 2 0 3 pellets or with B 4 C rods replacing fuel rods. The presence of these Gd rods and B 4 C rods is not credited in the analysis.
uncertainty value increased the total uncertainty by about 0.01 ilk.
This represents significant unquantified margin. From a generic study presented in Reference 15, not modeling the Gd 2 0 3 represents margin ranging from about 0.08 L\k at zero burnup to 0.002 L\k at high burnups. From Figure 36 in the Reference 15, not modeling the B 4 C rods in a CE 14x14 assembly results in margin ranging from around 0.15 L\k at low burnups to 0.002 L\k at burnups around 30 GWdlMTU. OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETl\RY INFORMATION An uncertainty equal to [ ] of the minor actinide and fission product worth was used Recent work documented in Reference 8 indicates that use of an uncertainty as low as 1.S percent of the minor actinide and fission product worth may be appropriate.
3.3.S.3 Conclusion on Analysis of Margins Considering both the potential nonconservatisms identified in Section 3.3.S.1 and the conservatisms identified in Section 3.3.S.2, it is concluded that the available margins offset the potential nonconservatisms.
Use of the [ ] uncertainty adds margin to the uncertainty analysis.
3.4     The NFV NCS Analysis During the review of the NCS analysis of the SFP it was determined that NCS analysis supporting fresh fuel in the NFV was needed to support related TS changes. Specifically, the TS limiting the fuel stored to a maximum uranium-23S enrichment of 4.S weight percent was changed to permit a maximum lattice average uranium-23S enrichment of 4.6 weight-percent uranium-23S. A NFV NCS analysis, documented in Holtec International report HI-2094416, Revision 1 (Reference 16). was provided as attachment 4 to Reference 6.
The "[ ] of the worth" uncertainty results in uncertainties ranging from 0.00S4 ilk at low burnups to 0.026 ilk at high burnups. At high burnups, use of the [ ] uncertainty value increased the total uncertainty by about 0.01 ilk. 3.3.S.3 Conclusion on Analysis of Margins Considering both the potential nonconservatisms identified in Section 3.3.S.1 and the conservatisms identified in Section 3.3.S.2, it is concluded that the available margins offset the potential nonconservatisms.
This section documents the review of the NFV NCS analysis.
3.4 The NFV NCS Analysis During the review of the NCS analysis of the SFP it was determined that NCS analysis supporting fresh fuel in the NFV was needed to support related TS changes. Specifically, the TS limiting the fuel stored to a maximum uranium-23S enrichment of 4.S weight percent was changed to permit a maximum lattice average uranium-23S enrichment of 4.6 weight-percent uranium-23S.
3.4.1   New Fuel Vault (NFV) NCS Analysis Method The analysis is to demonstrate compliance with the following requirements from 10 CFR SO.68:
A NFV NCS analysis, documented in Holtec International report HI-2094416, Revision 1 (Reference 16). was provided as attachment 4 to Reference
Paragraph SO.68(b)(2) of 10 CFR requires, "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 0.9S, at a 9S-percent probability, 9S-percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used."
: 6. This section documents the review of the NFV NCS analysis.
Paragraph SO.68(b)(3) of 10 CFR requires, "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 9S-percent probability. 9S-percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used."
3.4.1 New Fuel Vault (NFV) NCS Analysis Method The analysis is to demonstrate compliance with the following requirements from 10 CFR SO.68: Paragraph SO.68(b)(2) of 10 CFR requires, "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 0.9S, at a 9S-percent probability, 9S-percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used." Paragraph SO.68(b)(3) of 10 CFR requires, "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 9S-percent probability.
Compliance with these requirements is demonstrated in HI-2094416 by modeling a simplified representation of fresh fuel stored in the NFV submitted by the licensee as part of its application. The licensee's analysis demonstrates that the keff value, including bias and uncertainties, for both full density water and optimum moderation conditions is more than 0.03 ilk below the applicable limits from 10 CFR SO.68.
9S-percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used." Compliance with these requirements is demonstrated in HI-2094416 by modeling a simplified representation of fresh fuel stored in the NFV submitted by the licensee as part of its application.
OFFICIAL USE ONLY         PROPRIETARY INFORMATION
The licensee's analysis demonstrates that the keff value, including bias and uncertainties, for both full density water and optimum moderation conditions is more than 0.03 ilk below the applicable limits from 10 CFR SO.68. OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION  
 
-14 The MCNP4a computer code and nuclear data was used in their analysis to calculate the keff value for fresh fuel in the NFV fuel storage racks. MCNP4a has a long history of use for this type of analysis and is, therefore, acceptable.
OFFICIAL USE ONLY       PROPRIETARY INFORMATION
MCNP4a and its associated nuclear data were validated in Holtec International report HI-2094486RO, "MCNP Benchmark Calculations," and Section 2 of HI-2094416 states the bias and 95/95 uncertainty from the validation are 0.0013 ilk +/- 0.0086. This bias and bias uncertainty are similar to values reported for other analyses.
                                                - 14 The MCNP4a computer code and nuclear data was used in their analysis to calculate the keff value for fresh fuel in the NFV fuel storage racks. MCNP4a has a long history of use for this type of analysis and is, therefore, acceptable.
ConSidering that the values are consistent with similar analyses and that there is more than 0.03 ilk margin to the limits, further review of the validation was deemed unnecessary.
MCNP4a and its associated nuclear data were validated in Holtec International report HI-2094486RO, "MCNP Benchmark Calculations," and Section 2 of HI-2094416 states the bias and 95/95 uncertainty from the validation are 0.0013 ilk +/- 0.0086. This bias and bias uncertainty are similar to values reported for other analyses. ConSidering that the values are consistent with similar analyses and that there is more than 0.03 ilk margin to the limits, further review of the validation was deemed unnecessary.
The NRC staff reviewed the computational method and supporting validation and finds them acceptable.
The NRC staff reviewed the computational method and supporting validation and finds them acceptable.
3.4.2 NFV Fuel Storage Racks The steel structures that comprise the NFV fuel storage racks were conservatively not modeled. Without the steel, the rack model simplifies down to constraints on the spacing and location of the fuel assemblies.
3.4.2   NFV Fuel Storage Racks The steel structures that comprise the NFV fuel storage racks were conservatively not modeled.
All rack structures were modeled as water at the calculation-specific water density. Section 7 of HI-2094416 states that rack tolerances are not included since the racks are not modeled. Rack tolerances include tolerances on spacing between assemblies and between assemblies and the NFV floor and walls. Not modeling the storage rack steel provides enough margin to cover deficiencies in the uncertainty analysis.
Without the steel, the rack model simplifies down to constraints on the spacing and location of the fuel assemblies. All rack structures were modeled as water at the calculation-specific water density.
3.4.3 Fuel Assemblies HI-2094416 states that the design basis fuel assembly is a Westinghouse 16x16 and describes the assembly in Table 5.1 of that report. From Table 5.1, the 16x16 assembly is equivalent to a CE 16x16 fuel assembly.
Section 7 of HI-2094416 states that rack tolerances are not included since the racks are not modeled. Rack tolerances include tolerances on spacing between assemblies and between assemblies and the NFV floor and walls. Not modeling the storage rack steel provides enough margin to cover deficiencies in the uncertainty analysis.
The assembly used in the analysis presented in HI-2094416 is consistent with the fuel assembly design used for the spent fuel storage analysis documented in HI-2104753, review of which was covered in Section 3.3 of this safety evaluation.
3.4.3   Fuel Assemblies HI-2094416 states that the design basis fuel assembly is a Westinghouse 16x16 and describes the assembly in Table 5.1 of that report. From Table 5.1, the 16x16 assembly is equivalent to a CE 16x16 fuel assembly. The assembly used in the analysis presented in HI-2094416 is consistent with the fuel assembly design used for the spent fuel storage analysis documented in HI-2104753, review of which was covered in Section 3.3 of this safety evaluation.
The fuel assembly design is conservative in that natural uranium blankets and fuel rods containing Gd 2 0 3 are modeled as un poisoned 4.6 weight-percent uranium-235 U0 2. The analysis did not address the impact of using the planar average enrichment simplification.
The fuel assembly design is conservative in that natural uranium blankets and fuel rods containing Gd2 0 3 are modeled as unpoisoned 4.6 weight-percent uranium-235 U0 2. The analysis did not address the impact of using the planar average enrichment simplification. As was shown in Table 7.34 of HI-2104753, the impact of using average enrichments rather than distributed enrichments is small and mixed in direction. Any nonconservatism related to this effect can be covered by the large margins to the limits.
As was shown in Table 7.34 of HI-2104753, the impact of using average enrichments rather than distributed enrichments is small and mixed in direction.
Other than the active lengths of the fuel rods, other fuel assembly components were replaced with water. This is conservative for NVF analysis.
Any nonconservatism related to this effect can be covered by the large margins to the limits. Other than the active lengths of the fuel rods, other fuel assembly components were replaced with water. This is conservative for NVF analysis. Based on the NRC staffs review of the above, the fuel assembly model used in the NFV NCS analysis is adequately conservative.
Based on the NRC staffs review of the above, the fuel assembly model used in the NFV NCS analysis is adequately conservative.
OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY  
OFFICIAL USE ONLY       PROPRIETARY INFORMATION
-3.4.4 Analysis of Margins The licensee determined that the maximum keff value, including biases and uncertainties, for the full density flooding case was 0.9152. The applicable limit is 0.95. The margin to this limit is 0.035 .1.k. The licensee determined that the maximum keff value, including biases and uncertainties, for the optimum moderation case was 0.9358. The applicable limit is 0.98. Thus, the margin to this limit is 0.044 .1.k and the NRC staff finds this acceptable  
 
.. A few small potential nonconservatisms were identified in the review. The margins to the limits are large enough to cover potential nonconservatisms associated with deficiencies in the analysis of tolerances and failure to evaluate the impact of some modeling simplifications.  
OFFICIAL USE ONLY         PROPRIETARY INFORMATION
                                                - 15 3.4.4   Analysis of Margins The licensee determined that the maximum keff value, including biases and uncertainties, for the full density flooding case was 0.9152. The applicable limit is 0.95. The margin to this limit is 0.035 .1.k. The licensee determined that the maximum keff value, including biases and uncertainties, for the optimum moderation case was 0.9358. The applicable limit is 0.98. Thus, the margin to this limit is 0.044 .1.k and the NRC staff finds this acceptable ..
A few small potential nonconservatisms were identified in the review. The margins to the limits are large enough to cover potential nonconservatisms associated with deficiencies in the analysis of tolerances and failure to evaluate the impact of some modeling simplifications.
3.5      Conclusions The NRC staff review of the St. Lucie 2 spent fuel storage racks NCS analysis, documented in HI-2104753, and of the St. Lucie 2 NFV NCS analysis documented in HI-2094416, Revision 1, identified some nonconservative items.
Those items were evaluated against the margin to the regulatory limit and what the NRC considers an appropriate amount of margin attributable to conservatisms documented in the analyses.
It was noted that the results from nonconverged Monte Carlo calculations were used in the analysis. The response to RAI SRXB-121 indicated that upon rerunning many of the cases, the reported maximum keff values may be non conservative by as much as 0.005 .1.k. The NRC staff finds this acceptable because the burnup credit loading curves were designed to have a maximum keff' including biases and uncertainties, of 0.99. This left adequate margin to cover the use of results from poorly converged Monte Carlo calculations.
The uncertainty analysis included an allowance of only 2.5 percent of the assembly average burnup used to compare to the burnup credit loading curves. This value is lower than the more typical value used of 5 percent of assembly average burn up.
The effects of keff variation with SFP water temperature and density were quantified only at the extremes of the parameter ranges. Because the analysis did not include intermediate water densities to show that keff did not peak between the extremes of the parameter ranges, the non conservatism associated with this part of the analysis is likely between 0 and 0.0034 .1.k.
The licensee's use of a method for evaluating manufacturing tolerances and uncertainties in its SFP analysis was supported with calculations demonstrating that the revised method yielded conservative total uncertainties with additional margins ranging from 0.0014 to 0.0163 .1.k.
Following review of the supporting analysis reports and based on the margins to regulatory limits, crediting some analysis conservatism, and including consideration of the identified potential nonconservatisms, the NRC staff concludes that there is a reasonable assurance that the St. Lucie 2 SFP and NFV fuel storage racks meet the applicable NCS regulatory requirements.
OFFICIAL USE ONLY        PROPRIETARY INFORMP.TION
 
OFFICIAL USE ONLY      PROPRIETARY INFORMATION
                                                - 16 Because of the need to evaluate offsetting effects in the licensee's analysis, its analysis constitutes a methodology of which any change would be a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.
 
==4.0      STATE CONSULTATION==


===3.5 Conclusions===
Based upon a letter dated May 2, 2003, from Michael N. Stephens of the Florida Department of Health, Bureau of Radiation Control, to Brenda L. Mozafari, Senior Project Manager, U.S. Nuclear Regulatory Commission, the State of Florida does not desire notification of issuance of license amendments.


The NRC staff review of the St. Lucie 2 spent fuel storage racks NCS analysis, documented in HI-2104753, and of the St. Lucie 2 NFV NCS analysis documented in HI-2094416, Revision 1, identified some nonconservative items. Those items were evaluated against the margin to the regulatory limit and what the NRC considers an appropriate amount of margin attributable to conservatisms documented in the analyses.
==5.0     ENVIRONMENTAL CONSIDERATION==
It was noted that the results from nonconverged Monte Carlo calculations were used in the analysis.
The response to RAI SRXB-121 indicated that upon rerunning many of the cases, the reported maximum keff values may be non conservative by as much as 0.005 .1.k. The NRC staff finds this acceptable because the burnup credit loading curves were designed to have a maximum keff' including biases and uncertainties, of 0.99. This left adequate margin to cover the use of results from poorly converged Monte Carlo calculations.
The uncertainty analysis included an allowance of only 2.5 percent of the assembly average burnup used to compare to the burnup credit loading curves. This value is lower than the more typical value used of 5 percent of assembly average burn up. The effects of keff variation with SFP water temperature and density were quantified only at the extremes of the parameter ranges. Because the analysis did not include intermediate water densities to show that keff did not peak between the extremes of the parameter ranges, the non conservatism associated with this part of the analysis is likely between 0 and 0.0034 .1.k. The licensee's use of a method for evaluating manufacturing tolerances and uncertainties in its SFP analysis was supported with calculations demonstrating that the revised method yielded conservative total uncertainties with additional margins ranging from 0.0014 to 0.0163 .1.k. Following review of the supporting analysis reports and based on the margins to regulatory limits, crediting some analysis conservatism, and including consideration of the identified potential nonconservatisms, the NRC staff concludes that there is a reasonable assurance that the St. Lucie 2 SFP and NFV fuel storage racks meet the applicable NCS regulatory requirements.
OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION
-16 Because of the need to evaluate offsetting effects in the licensee's analysis, its analysis constitutes a methodology of which any change would be a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.


==4.0 STATE CONSULTATION==
Pursuant to 10 CFR 51.21, 51.32, 51.33, and 51.35, a draft Environmental Assessment and finding of no significant impact was prepared and published in the Federal Register on January 6,2012 (77 FR 813). The draft Environmental Assessment provided a 30-day opportunity for public comment. The NRC staff received comments that were addressed in the final environmental assessment. The final Environmental Assessment was published in the Federal Register on July 6,2012 (77 FR 40092). Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.


Based upon a letter dated May 2, 2003, from Michael N. Stephens of the Florida Department of Health, Bureau of Radiation Control, to Brenda L. Mozafari, Senior Project Manager, U.S. Nuclear Regulatory Commission, the State of Florida does not desire notification of issuance of license amendments.  
==6.0      CONCLUSION==


===5.0 ENVIRONMENTAL===
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) such activities will be conducted in compliance with the Commission's 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.


CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, 51.33, and 51.35, a draft Environmental Assessment and finding of no significant impact was prepared and published in the Federal Register on January 6,2012 (77 FR 813). The draft Environmental Assessment provided a 30-day opportunity for public comment. The NRC staff received comments that were addressed in the final environmental assessment.
==7.0      REFERENCES==
The final Environmental Assessment was published in the Federal Register on July 6,2012 (77 FR 40092). Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.
: 1. Florida Power and Light Company letter L-2011-021, RL. Anderson, Site Vice President, St. Lucie Plant to USNRC document control desk, re: "St. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, License Amendment Request for Extended Power Uprate, n February 25, 2011 (ADAMS ML110730116).
: 2. Florida Power and Light Company letter L-2011-466, RL. Anderson, Site Vice President, St. Lucie Plant to USNRC document control desk, re: "St. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Revision to Extended Power Uprate License Amendment Request Proposed Technical Specification 5.6, Design Features - Fuel Storage - Criticality," November 4,2011 (ADAMS ML11314A111).
: 3. Florida Power and Light Company letter L-2012-181, RL. Anderson, Site Vice President, St. Lucie Plant to USNRC document control desk, re: "St. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Extended Power Uprate License Amendment Request - Supplement to Proposed Technical Specification Changes Related to Spent Fuel Storage Requirements," April 30, 2012 (ADAMS ML12124A065).
OFFICIAL USE ONLY        PROPRIETARY INFORMATION


==6.0 CONCLUSION==
OFFICIAL USE ONLY        PROPRIETARY INFORMATION
                                                - 17
: 4. Florida Power and Light Company letter L-2011-489, R L. Anderson, Site Vice President, S1. Lucie Plant to USNRC document control desk, re: "S1. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment," December 8, 2011 (ADAMS ML11350A245).
: 5. Florida Power and Light Company letter L-2012-182, RL. Anderson, Site Vice President, S1. Lucie Plant to USNRC document control desk, re: "S1. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request," May 4,2012 (ADAMS ML12130A478).
: 6. Florida Power and Light Company letter L-2012-201, RL. Anderson, Site Vice President, S1. Lucie Plant to USNRC document control desk, re: "S1. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request," May 7,2012 (ADAMS ML12132A414).
: 7. S. Anton, "Lumped Fission Products and Pm148m Cross Sections for MCNP," HI-2033031, Holtec International, January 2011.
: 8. J.M. Scaglione, D.E. Mueller, J.C. Wagner, and W.J. Marshall, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses - Criticality (kef!)
Predictions," NUREG/CR-7109 (ORNLrrM-2011/514), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, April 2012.
: 9. NRC Memorandum from L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants,"
August 19, 1998 (ADAMS ML003728001).
: 10. V. Makodym, "Nuclear Group Computer Code Benchmark Calculations," HI-2104790, Holtec International, January 2011 (ADAMS ML11350A245).
: 11. D.E. Mueller, K.R Elam, and P.B. Fox, "Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data," NUREG/CR-6979, ORNLfTM-2007/083, U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, September 2008.
: 12. J.C. Wagner, M.D. DeHart, and C.V. Parks, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis," NUREG/CR-6801 (ORNLfTM-2001/273), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2003.
: 13. C. Kang, et aI., "Horizontal Burnup Gradient Datafile for PWR Assemblies," DOE/RW-0496, U.S. Department of Energy, Office of Civilian Radioactive Waste Management, May 1997.
: 14. J.C. Wagner and C.E. Sanders, "Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs," NUREG/CR-6800 (ORNLrrM-2002/6), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2003.
OFFICIAL USE ONLY        PROPRIETARY INFORMATION


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) such activities will be conducted in compliance with the Commission's 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.  
OFFICIAL USE ONLY          PROPRIETARY INFORMATION
                                                  - 18
: 15. CV. Parks, M.D. DeHart, and J.C. Wagner, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," NUREG/CR-6665 (ORNLlTM-1999/303),
U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, February 2000.
: 16. B. Brickner, "Criticality Analysis of the St. Lucie Unit 2 New Fuel Vault," HI-2094416, Revision 1, Holtec International, June 2010.
Principal Contributor: Kent Wood Date: September 19, 2012 OFFICIAL USE ONLY          PROPRIETARY INFORMATION


==7.0 REFERENCES==
OFFICIAL USE ONLY         PROPRIETARY INFORMATION M. Nazar                                           -2 The NRC has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations, Section 2.390, "Public Inspections, Exemptions, Requests for Withholding." Accordingly, the NRC staff has also prepared a redacted, publicly-available, non-proprietary version of the SE. Copies of the proprietary and non-proprietary versions of the SE are enclosed.
Florida Power and Light Company letter L-2011-021, RL. Anderson, Site Vice President, St. Lucie Plant to USNRC document control desk, re: "St. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, License Amendment Request for Extended Power Uprate, n February 25, 2011 (ADAMS ML 110730116). Florida Power and Light Company letter L-2011-466, RL. Anderson, Site Vice President, St. Lucie Plant to USNRC document control desk, re: "St. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Revision to Extended Power Uprate License Amendment Request Proposed Technical Specification 5.6, Design Features -Fuel Storage -Criticality," November 4,2011 (ADAMS ML 11314A111). Florida Power and Light Company letter L-2012-181, RL. Anderson, Site Vice President, St. Lucie Plant to USNRC document control desk, re: "St. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Extended Power Uprate License Amendment Request -Supplement to Proposed Technical Specification Changes Related to Spent Fuel Storage Requirements," April 30, 2012 (ADAMS ML 12124A065).
A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY
Sincerely, IRA!
-17 Florida Power and Light Company letter L-2011-489, R L. Anderson, Site Vice President, S1. Lucie Plant to USNRC document control desk, re: "S1. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment," December 8, 2011 (ADAMS ML 11350A245). Florida Power and Light Company letter L-2012-182, RL. Anderson, Site Vice President, S1. Lucie Plant to USNRC document control desk, re: "S1. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request," May 4,2012 (ADAMS ML 12130A478). Florida Power and Light Company letter L-2012-201, RL. Anderson, Site Vice President, S1. Lucie Plant to USNRC document control desk, re: "S1. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request," May 7,2012 (ADAMS ML 12132A414). S. Anton, "Lumped Fission Products and Pm148m Cross Sections for MCNP," HI-2033031, Holtec International, January 2011. J.M. Scaglione, D.E. Mueller, J.C. Wagner, and W.J. Marshall, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses -Criticality (kef!) Predictions," NUREG/CR-7109 (ORNLrrM-2011/514), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, April 2012. NRC Memorandum from L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," August 19, 1998 (ADAMS ML003728001).
Tracy J. Ort, Project Manager Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389
: 10. V. Makodym, "Nuclear Group Computer Code Benchmark Calculations," HI-2104790, Holtec International, January 2011 (ADAMS ML 11350A245).
: 11. D.E. Mueller, K.R Elam, and P.B. Fox, "Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data," NUREG/CR-6979, ORNLfTM-2007/083, U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, September 2008. 12. J.C. Wagner, M.D. DeHart, and C.V. Parks, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis," NUREG/CR-6801 (ORNLfTM-2001/273), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2003. 13. C. Kang, et aI., "Horizontal Burnup Gradient Datafile for PWR Assemblies," DOE/RW-0496, U.S. Department of Energy, Office of Civilian Radioactive Waste Management, May 1997. 14. J.C. Wagner and C.E. Sanders, "Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs," NUREG/CR-6800 (ORNLrrM-2002/6), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2003. OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION
-18 15. CV. Parks, M.D. DeHart, and J.C. Wagner, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," NUREG/CR-6665 (ORNLlTM-1999/303), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, February 2000. 16. B. Brickner, "Criticality Analysis of the St. Lucie Unit 2 New Fuel Vault," HI-2094416, Revision 1, Holtec International, June 2010. Principal Contributor:
Kent Wood Date: September 19, 2012 OFFICIAL USE ONLY PROPRIETARY OFFICIAL USE ONLY PROPRIETARY INFORMATION M. Nazar -2 The NRC has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations, Section 2.390, "Public Inspections, Exemptions, Requests for Withholding." Accordingly, the NRC staff has also prepared a redacted, publicly-available, non-proprietary version of the SE. Copies of the proprietary and non-proprietary versions of the SE are enclosed.
A copy of the Safety Evaluation is also enclosed.
The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, IRA! Tracy J. Ort, Project Manager Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389  


==Enclosures:==
==Enclosures:==
: 1. Amendment No. 162 to NPF-16 2. Non-Proprietary Safety Evaluation
: 1. Amendment No. 162 to NPF-16
: 2. Non-Proprietary Safety Evaluation
: 3. Proprietary Safety Evaluation cc wI Enclosures 1 and 2: Distribution via Listserv Distribution:
: 3. Proprietary Safety Evaluation cc wI Enclosures 1 and 2: Distribution via Listserv Distribution:
PUBLIC RidsNrrDorlDpr LPL2-2 rtf RidsNrrDssStsb RidsNrrDorlLpl2-2 RidsOgcRp RidsNrrLABClayton RidsRgn2MailCenter RidsNrrPMStLucie RidsAcrsAcnw
PUBLIC                                                     RidsNrrDorlDpr LPL2-2 rtf                                                 RidsNrrDssStsb RidsNrrDorlLpl2-2                                           RidsOgcRp RidsNrrLABClayton                                           RidsRgn2MailCenter RidsNrrPMStLucie                                           RidsAcrsAcnw_MaiICTR RidsNrrDssSrxb                                             KWood, NRR ADAMS Accession No. Non-Proprietary SE) ML12263A224            (Proprietary SE ML12243A415 OFFICE     LPL2-2/PM   LPL2-2/LA   SRXB/BC       STSB/BC         OGC     NLO LPL2-2/BC (A) LPL2-2/PM NAME       TOrt         BClayton     CJackson     RElliott       LSubin       JQuichocho   TOrt DATE         9118/2012   9/17/2012   9/5/2012     9/12/2012       9/1012012     9/18/2012   911912012 OFFICIAL RECORD COpy OFFICIAL USE ONLY         PROPRIETARY INFORMATION}}
_MaiICTR RidsNrrDssSrxb KWood, NRR ADAMS Accession No. Non-Proprietary SE) ML 12263A224 (Proprietary SE ML 12243A415 OFFICE LPL2-2/PM LPL2-2/LA SRXB/BC STSB/BC OGC NLO LPL2-2/BC (A) LPL2-2/PM NAME TOrt BClayton CJackson RElliott LSubin JQuichocho TOrt DATE 9118/2012 9/17/2012 9/5/2012 9/12/2012 9/1012012 9/18/2012 911912012 OFFICIAL RECORD OFFICIAL USE ONLY PROPRIETARY}}

Latest revision as of 13:09, 6 February 2020

Public - Issuance of Amendment Regarding New Fuel Vault and Spent Fuel Pool Nuclear Criticality Analysis
ML12263A224
Person / Time
Site: Saint Lucie NextEra Energy icon.png
Issue date: 09/19/2012
From: Orf T
Plant Licensing Branch II
To: Nazar M
Florida Power & Light Co
Orf, T J
References
TAC ME8782
Download: ML12263A224 (39)
v • d • e


Text

OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 19, 2012 Mr. Mano Nazar Executive Vice President and Chief Nuclear Officer Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420

SUBJECT:

ST. LUCIE PLANT, UNIT 2 - ISSUANCE OF AMENDMENT REGARDING NEW FUEL VAULT AND SPENT FUEL POOL NUCLEAR CRITICALITY ANALYSIS (TAC NO. ME8782)

Dear Mr. Nazar:

The Commission has issued the enclosed Amendment No. 162 to Renewed Facility Operating License No. NPF-16 for the st. Lucie Plant, Unit No.2. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated February 25, 2011, as supplemented by letters dated November 4 and December 8, 2011, and April 30 and May 4 and 7, 2012.

This amendment raises the maximum fuel enrichment for fresh fuel storage from a maximum of 4.5 weight-percent uranium-235 to a maximum lattice averaged value of 4.6 weight-percent uranium-235. The TS changes associated with fuel stored in the spent fuel pool include increasing the maximum initial enrichment from 4.5 weight-percent uranium-235 to a maximum planar average initial enrichment of 4.6 weight-percent uranium-235, credit for empty storage locations, credit for use of METAMICTM inserts, credit for installation of full-length full-strength, five-fingered control element assemblies, and definition of three special configurations referred to in the nuclear criticality safety analysis as inspection and maintenance configurations.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORM,AJ"ION M. Nazar -2 The NRC has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations, Section 2.390, "Public Inspections, Exemptions, Requests for Withholding." Accordingly, the NRC staff has also prepared a redacted, publicly-available, non-proprietary version of the SE. Copies of theproprietary and non-proprietary versions of the SE are enclosed.

A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Tracy J. Ort, Project Manager Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389

Enclosures:

1. Amendment No. 162 to NPF-16
2. Non-Proprietary Safety Evaluation
3. Proprietary Safety Evaluation cc wI Enclosures 1 and 2: Distribution via Listserv OFFICIAL USE ONLY PROPRIETARY INFORM,A,TION

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FLORIDA POWER & LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO, FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST. LUCIE PLANT UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 162 Renewed License No. NPF-16

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Florida Power & Light Company (the licensee), dated February 25,2011, as supplemented by letters dated November 4 and December 8,2011, and April 30 and May 4 and 7, 2012; complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

-2

2. Accordingly, Renewed Facility Operating License No. NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 3.B to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 162, are hereby incorporated in the renewed license.

FPL shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

-'7J7_~'~~

Ie . Quichocho, Acting Chief ant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: September 18, 2012

ATTACHMENT TO LICENSE AMENDMENT NO. 162 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 Replace Page 3 of Renewed Operating License NPF-16 with the attached Page 3.

Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages XXII XXII XXV XXV 3/49-12 3/49-12 5-4 5-4 5-4a 5-4a 5-4b 5-4b 5-4c 5-4d 5-4e 5-4f 5-4g 5-4h 5-4i 5-4j 5-4k 5-41 5-4m 5-4n 5-40

-3 neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required.

D. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission's regulations: 10 CFR Part 20, Section 30.34 of 10 FR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all 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:

A. Maximum Power Level FPL is authorized to operate the facility at steady state reactor core power levels not in excess of 2700 megawatts (thermal).

Commencing with the startup for Cycle 16 and until the Combustion Engineering Model 3410 Steam Generators are replaced, the maximum reactor core power shall not exceed 89 percent of 2700 megawatts (thermal) if:

a. The Reactor Coolant System Flow Rate is less than 335,000 gpm but greater than or equal to 300,000 gpm, or
b. The Reactor Coolant System Flow Rate is greater than or equal to 300,000 gpm AND the percentage of steam generator tubes plugged is greater than 30 percent (2520 tubes/SG) but less than or equal to 42 percent (3532 tubes/SG).

This restriction in maximum reactor core power is based on analyses provided by FPL in submittals dated October 21, 2005 and February 28, 2006, and approved by the NRC in Amendment No. 145, which limits the percent of steam generator tubes plugged to a maximum of 42 percent (3532 tubes) in either steam generator and limits the plugging asymmetry between steam generators to a maximum of 600 tubes.

B. Technical SpeCifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 162 are hereby incorporated in the renewed license.

FPL shall operate the facility in accordance with the Technical Specifications.

Renewed License No. NPF-16 Amendment No. 162

INDEX LIST OF FIGURES (continued)

FIGURE PAGE 3.4-3 ST. LUCIE UNIT 2 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS FOR 55 EFPY, COOLDOWN AND INSERVICE TEST.............................................................................................................3/4 4-31b 3.4-4 DELETED .......................................................................................................3/4 4-32 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST .............................. 3/4 7-25 5.1-1 SITE AREA MAP....................................................................................................5-2 5.6-1a DELETED .....................................................................................................................

5.6-1 b DELETED....................................................................................................................

5.6-1c DELETED ....................................................................................................................

5.6-1d DELETED ....................................................................................................................

5.6-1e DELETED ....................................................................................................................

5.6-1f DELETED ....................................................................................................................

5.6-1 ALLOWABLE REGION 1 STORAGE PATTERNS AND FUEL ARRANGEMENTS ...............................................................................................5-4h 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 1 of 3).........................................................................5-4i 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 2 of 3) .........................................................................5-4j 5.6-2 ALLOWABLE REGION 2 STORAGE PATTERNS AND FUEL ARRANGEMENTS (Sheet 3 of 3)........................................................................5-4k 5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 1 of 2) .........................................................................................................5-41 5.6-3 INTERFACE REQUIREMENTS BETWEEN REGION 1 AND REGION 2 (Sheet 2 of 2).......................................................................................................5-4m 5.6-4 ALLOWABLE CASK PIT STORAGE RACK PATTERNS .................................... 5-4n 6.2-1 DELETED ...............................................................................................................6-3 6.2-2 DELETED...............................................................................................................6-4 ST. LUCtE - UNIT 2 XXII Amendment No. 8, ~, 56, ~, 442.

-147. 43i,4a4, 162

INDEX LIST OF TABLES (Continued)

TABLE 3.7-1 MAXIMUM ALLOWABLE LINEAR POWER LEVEL-HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFElY VALVES DURING OPERA TION WITH BOTH STEAM GENERATORS ............................................................ 3/47-2 3.7-2 STEAM LINE SAFElY VALVES PER LOOP .......................................................... 3/4 7-3 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVllY SAMPLE AND ANALYSIS PROGRAM ............................................................................................ 3/4 7-8 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL ...................................................... 314 7-22 3.7-3a DELETED ............................................................................................................... 3/4 7-26 3.7-3b DELETED ...............................................................................................................3/47-27 3.7-4 DELETED 3.7-5 DELETED 4.8-1 DIESEL GENERATOR TEST SCHEDULE .............................................................. 3148-8 4.8-2 BATTERY SURVEILLANCE REQUIREMENT ...................................................... 3/ 48-12 3.8-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES ................................................................................................ 3/48-18 4.11-1 DELETED 4.11-2 DELETED 3.12-1 DELETED 3.12-2 DELETED 4.12-1 DELETED 5.6-1 MINIMUM BURNUP COEFFiCiENTS .......................................................................... 5-40 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS ......................................................... 5-5 6.2-1 MINIMUM SHIFT CREW COMPOSITION-TWO UNITS WITH TWO SEPARATE CONTROL ROOMS ...................................................................................6-5 ST. LUCIE - UNIT 2 Amendment No. S, 84. 93. 68.

+3,..f.i7.,162

REFUELING OPERATIONS 3/4.9.11 SPENT FUEL STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 The Spent Fuel Pool shall be maintained with:

a. The fuel storage pool water level greater than or equal to 23 ft over the top of irradiated fuel assemblies seated in the storage racks, and
b. The fuel storage pool boron concentration greater than or equal to 1900 ppm.

APPLICABILITY: Whenever irradiated fuel assemblies are in the spent fuel storage pool.

ACTION:

a. With the water level requirement not satisfied, immediately suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b. With the boron concentration requirement not satisfied, immediately suspend all movement of fuel assemblies in the fuel storage pool and initiate action to restore fuel storage pool boron concentration to within the required limit.
c. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the spent fuel storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.

4.9.11.1 Verify the fuel storage pool boron concentration is within limit at least once per 7 days.

ST. LUCIE - UNIT 2 3/49-12 Amendment No. 4M, 162

DESlGN FEATURES 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

5.6 FUEL STORAGE CRITICALITY 5.6.1 a. The spent fuel storage racks are designed and shall be maintained with:

1. A keff equivalent to less than 1.0 when flooded with unborated water, including a conservative allowance for biases and uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.
2. A keff equivalent to less than or equal to 0.95 when flooded with water containing 500 ppm boron, including a conservative allowance for biases and uncertainties as described in Section 9.1 of the Updated Final Safety Analysis Report.
3. A nominalS.965 inch center-to-center distance between fuel assemblies placed in the spent fuel pool storage racks and a nominal S.SO inch center to-center distance between fuel assemblies placed in the cask pit storage rack.
4. For storage of enriched fuel assemblies, requirements of SpeCification 5.6.1.a.1 and 5.6.1.a.2 shall be met by positioning fuel in the spent fuel pool storage racks consistent with the requirements of Specification 5.6.1.c.
5. Fissile material, not contained in a fuel assembly lattice, shall be stored in accordance with the requirements of Specifications 5.6.1.a.1 and 5.6.1.a.2.
6. The Metamic neutron absorber inserts shall have a 1°B areal density greater than or equal to 0.015 grams 10B/cm2.
b. The cask pit storage rack shall contain neutron absorbing material (Boral) between stored fuel assemblies when installed in the spent fuel pool.
c. Loading of spent fuel pool storage racks shall be controlled as described below.
1. The maximum initial planar average U-235 enrichment of any fuel assembly inserted in a spent fuel pool storage rack shall be less than or equal to 4.6 weight percent.
2. Fuel placed in Region 1 of the spent fuel pool storage racks shall comply with the storage pattern definitions of Figure 5.6-1 and the minimum bumup requirements as defined in Table 5.6-1. (See SpeCification 5.6.1.c.7 for exceptions)
3. Fuel placed in Region 2 of the spent fuel pool storage racks shall comply with the storage pattern definitions or allowed speCial arrangement definitions of Figure 5.6-2 and the minimum burnup requirements as defined in Table 5.6-1. (See Specification 5.6.1.c.7 for exceptions)

ST. LUCIE - UNIT 2 54 Amendment No.7, 9&,

162

  • . ~.~.

DESIGN FEATURES (continued)

CRITICALITY (continued) 5.6.1 c. (continued)

4. The 2x2 array of fuel assemblies that span the interface between Region 1 and Region 2 of the spent fuel pool storage racks shall comply with the storage pattern definitions of Figure 5.6-3 and the minimum burnup requirements as defined in Table 5.6-1. The allowed special arrangements in Region 2 as shown in Figure 5.6-2 shall not be placed adjacent to Region 1.

(See Specification 5.6.1.c.7 for exceptions)

5. Fuel placed in the cask pit storage rack shall comply with the storage pattern definitions of Figure 5.6-4 and the minimum burnup requirements as defined in Table 5.6-1. (See Specification 5.6.1.c.7 for exceptions)
6. The same directional orientation for Metamic inserts is required for contiguous groups of 2x2 arrays where Metamic inserts are required.
7. Fresh or spent fuel in any allowed configuration may be replaced with non fuel hardware, and fresh fuel in any allowed configuration may be replaced with a fuel rod storage basket containing fuel rod(s). Also, storage of Metamic inserts or control rods, without any fissile material, is acceptable in locations designated as completely water-filled cells.
d. The new fuel storage racks are designed for dry storage of unirradiated fuel assemblies having a maximum planar average U-235 enrichment less than or equal to 4.6 weight percent, while maintaining a keff of less than or equal to 0.98 under the most reactive condition.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 56 feet.

CAPACITY 5.6.3 The spent fuel pool storage racks are designed and shall be maintained with a storage capacity limited to no more than 1491 fuel assemblies, and the cask pit storage rack is designed and shall be maintained with a storage capacity limited to no more than 225 fuel assemblies. The total Unit 2 spent fuel pool and cask pit storage capacity is limited to no more than 1716 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

5T. LUCIE - UNIT 2 5-4a AmendmentNo.:f..*.m, 162 I

Pages 5-4C through 5-4F (Amendment 101) and page 5-4G (Amendment 135) have been deleted from the Technical Specifications.

The next page is 5-4h.

ST. LUCIE - UNIT 2 5-4b Amendment No. .f.M., 162

Allowable Storage Pattems (See Notes 1 and 2)

Pattem "A" Pattem "8" Pattem "Cft See Definition 1 See Definition 2 See Definition 3 c:JEJ [J[J [J[J DEFINITIONS:

EJc:J EJ[J EJ[J

1. Allowable pattem is fresh or bumed fuel checkerboarded with completely water-filled cells.

Diagram is for illustration only, where F represents Fuel and WH represents a completely water-filled cell.

2. Allowable pattern is placement of fuel assemblies that meet the requirements of type 1 in each 2x2 array location with at least one full-length full-strength CEA placed in any cell.

Minimum bumup for fuel assembly type 1 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

3. Allowable pattern is placement of fuel assemblies that meet the requirements of type 2 in three of the 2x2 array locations in combination with one completely water-filled cell.

Minimum burnup for fuel assembly type 2 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

NOTES:

1. The storage arrangements of fuel within a rack module may contain more than one pattern.

Each cell is a part of up to four 2x2 arrays, and each cell must simultaneously meet the requirements of all those arrays of which it is a part.

2. Completely water-filled cells within any pattern are acceptable.

FIGURE 5.6-1 Allowable Region 1 Storage Patterns and Fuel Arrangements ST. LUCIE - UNIT 2 5-4h Amendment No. 162

ALLOWED SPECIAL ARRANGEMENTS (See Notes 1 and 2)

Fresh Fuel Assemblies in Region 2 Racks See Definition SR 1 See Definition SR2 See Definition SR3

~ =

~

Fresh Fuel Assembly = Fuel Assembly with Metamic Insert I = Empty Cell D = Fuel Assembly per Definition ALLOWABLE STORAGE PAITERNS (See Notes 1 and 2)

Pattern uD" Pattern ME" Pattern "F" See Definition 1 See Definition 2 See Definition 3 Pattern "G" Pattern "H" See Definition 4 See Definition 5 FIGURE 5.6-2 (Sheet 1 of 3)

Allowable Region 2 Storage Patterns and Fuel AlTangements ST. LUCIE - UNIT 2 5-4i Amendment No. 162

DEFINITIONS for Figure 5.6*2

1. Allowable pattern is fuel assemblies that meet the requirements of type 3 in three of the 2x2 array locations in combination with one completely water-filled cell. Minimum burnup for fuel assembly type 3 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.
2. Allowable pattern is fuel assemblies that meet the requirements of type 4 in each of the 2x2 array locations with at least two Metamic inserts placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 4 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.
3. Allowable pattern is fuel assemblies that meet the requirements of type 5 in each of the 2x2 array locations with at least one Metamic insert placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 5 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.
4. Allowable pattern is fuel assemblies that meet the requirements of type 6 in each of the 2x2 array locations with at least two full-length. full strength 5 finger CEAs placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 6 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.
5. Allowable pattern is fuel assemblies that meet the requirements of type 7 in each of the 2x2 array locations with at least one full-length. full strength 5 finger CEA placed anywhere in the 2x2 array. Minimum burnup for fuel assembly type 7 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and coaling time. Diagram is for illustration only.

SR1. Allowable pattern is up to four fresh or bumed fuel assemblies placed in a 3x3 array in combination with Pattern "D" placed outside the 3x3 array. Fresh or burned fuel shall be placed in the comers of the 3x3 array with completely water-filled cells placed face-adjacent on all sides. A fuel assembly that meets the requirements of type 3 shall be placed in the center of the 3x3 array.

Minimum burnup for fuel assembly type 3 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

SR2. Allowable pattern is up to four fresh or burned fuel assemblies placed in a 3x3 array in combination with Pattern liE" placed outside the 3x3 array. Fresh or burned fuel shall be placed in the corners of the 3x3 array with completely water-tilled cells placed face-adJacent on all sides. A fuel assembly that meets the requirements of type 4 with a Metamic insert shall be placed in the center of the 3x3 array. Minimum burnup for fuel assembly type 4 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

FIGURE 5.6-2 (Sheet 2 of 3)

Allowable Region 2 Storage Patterns and Fuel Arrangements ST. LUCIE - UNrT 2 5-4j Amendment No. 162

SR3. Allowable pattern is up to four fresh or bumed fuel assemblies placed in a 3x3 array in combination with Pattem ifF" placed outside the 3x3 array. Fresh or bumed fuel shall be placed in the corners of the 3x3 array with completely water-filled cells placed face-adjacent on all sides. A fuel assembly that meets the requirements of type 5 with a Metamic insert shall be placed in the center of the 3x3 array. Minimum burnup for fuel assembly type 5 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment and cooling time. Diagram is for illustration only.

NOTES

1. The storage arrangements of fuel within a rack module may contain more than one pattern.

Each cell is a part of up to four 2x2 arrays, and each cell must simultaneously meet the requirements of all those arrays of which it is a part.

2. Completely water-filled cells within any pattem are acceptable.

FIGURE 5.6-2 (Sheet 3 of 3)

Allowable Region 2 Storage Patterns and Fuel Arrangements ST. LUCIE. - UNIT 2 5-4k Amendment No. 162

Allowed Region 2 to Region 1 Fuel Alignments (See Note 1)

Region 2 Interface of Region 2 Pattern "D" with Region 1 "---------------------------------------------------------------

See Definition 1

~EJ~EJ Region 1 Interface of Region 2 Pattern "E" with Region 1

~~~~ Region 2 See Definition 1

[J[J[J[J Region 1 Interface of Region 2 Pattern "F" with Region 1

~D~D Region 2 DODD See Definition 1 Region 1 Interface of Region 2 Pattern "Gil with Region 1 CJ~CJ~

Region 2 See Definition 1

~~~~ Region 1 Interface of Region 2 Pattern "H" with Region 1 DDDD Region 2 See Definition 1 Region 1 FIGURE 6.6-3 (Sheet 1 of 2)

Interface Requirements between Region 1 and Region 2 5-41 Amendment No. 162 ST. LUCIE - UNIT 2

DEFINITION:

1. Each 2x2 array that spans Region 1 and Region 2 shall match one of the Region 2 allowable storage patterns as defined by Specification S.6.1.c.3. Any required Metamic inserts must be placed into the fuel assemblies in Region 2. Locations of completely water-filled cells or CEAs may be in either Region 1 or Region 2. For interface assemblies, the requirements of Specifications S.6.1.c.2 and Specification S.6.1.c.3 shall be followed within Region 1 and Region 2, respectively. The Diagrams are for illustration only.

NOTES:

1. Completely water-filled cells within any pattern are acceptable.

FIGURE 5.6-3 (Sheet 2 of 2)

Interface Requirements between Region 1 and Region 2 ST. LUCIE - UNIT 2 5-4m Amendment No. 162

Allowable Storage Patterns (See Notes 1 and 2)

Pattern "'A" Pattern "B" See Definition 1 See Definition 2

[JEJ EJ[J DEFINITIONS:

1. Allowable pattern is fresh or burned fuel checkerboarded with completely water-filled cells.

Diagram is for illustration only, where F represents Fuel and WH represents a completely water-filled cell.

2. Allowable pattern is placement of fuel assemblies that meet the requirements of type 8 in three of the 2x2 array locations in combination with one completely water-filled cell in any location. Minimum burnup for fuel assembly type 8 is defined in Table 5.6-1 as a function of maximum initial planar average enrichment. Diagram is for illustration only.

NOTES:

1. The storage arrangements of fuel within a rack module may contain more than one pattern.

Each cell is a part of up to four 2x2 arrays. and each cel/ must simultaneously meet the requirements of all those arrays of which it is a part.

2. Completely water-filled cells within any pattern are acceptable.

FIGURE 5.6-4 Allowable Cask Pit Storage Rack Patterns ST. LUCIE - UNIT 2 5-4n Amendment No. 162

TABLE 5.6-1 Minimum Burnup Coefficients Cooling Time Coefficients Fuel Type (Years) A B C 1 0 -33.4237 25.6742 -1.6478 I

2 0 -25.3198 14.3200 -0.4042 i 3 0 -23.4150 16.2050 -0.5500 0 -33.2205 24.8136 -1.5199 I 2.5 -31.4959 23.4776 -1.4358 i 5 -30.4454 22.7456 -1.4147 4

10 -28.4361 21.2259 -1.2946 I 15 -27.2971 20.3746 -1.2333 20 -26.1673 19.4753 -1.1403 0 -24.8402 23.5991 -1.2082 2.5 -22.9981 21.6295 -1.0249 5 -21.8161 20.5067 -0.9440 5

10 -20.0864 19.0127 -0.8545 15 -19.4795 18.3741 -0.8318 20 -18.8225 17.7194 -0.7985 0 -32.4963 25.3143 -1.5534 2.5 -30.6688 23.6229 -1.4025 5 -29.2169 22.5424 -1.3274 i 6 10 -27.2539 21.0241 -1.2054 15 -25.7327 19.8655 -1.1091 20 -25.2717 19.5222 -1.1163 0 -24.6989 24.1660 -1.2578 2.5 -23.0399 22.3047 -1.0965 5 -21.2473 20.6553 -0.9403 7

10 -20.1775 19.5506 -0.9015 15 -19.4037 18.6626 -0.8490 20 -18.3326 17.7040 -0.7526 8 0 -43.4750 11.6250 0.0000 NOTES:

1. To qualify in a "fuel type", the burnup of a fuel assembly must exceed the minimum burnup "SU" calculated by inserting the "coefficients" for the associated "fuel type" and "cooling time" into the following polynomial function:

SU =A + S'*E + C*E2, where:

SU = Minimum Bumup (GWD/MTU)

E =Maximum Initial Planar Average Enrichment (weight percent U-235)

A, S, C =Coefficients for each fuel type

2. Interpolation between values of cooling time is not permitted.

ST. LUCIE* UNIT 2 5-40 Amendment No. 162

OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 162 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 FLORIDA POWER AND LIGHT COMPANY, ET AL.

ST. LUCIE PLANT. UNIT NO.2 DOCKET NOS. 50-389

1.0 INTRODUCTION

By letter dated February 25,2011 (Reference 1), Florida Power and Light Company (the licensee) requested to amend Renewed Facility Operating License No. NPF-16 and revise the St. Lucie, Unit No.2 (St. Lucie 2), Technical Specifications (TSs). The proposed amendment requested revisions to the Renewed Facility Operating License and TSs to support an extended power uprate (EPU) for St. Lucie, Unit 2, at a licensed core thermal power level increased from 2700 megawatts thermal (MWt) to 3020 MWt. For scheduling purposes, the review of the nuclear criticality safety (NCS) analysis for the St. Lucie 2 spent fuel pool (SFP) and new fuel vault (NFV) has been processed as a separate amendment request. The following evaluation presents the results of the U.S. Nuclear Regulatory Commission's (NRC) review of the NCS analysis.

The License Amendment Request (LAR) was amended by a letter dated November 4, 2011 (Reference 2). Attachment 3 to that letter included Revision 2 to Holtec International Report HI-21 04753. In subsequent submittals, the licensee provided revised proposed technical specifications (Reference 3); responses to NRC requests for additional information (RAls)

(References 4,5, and 6); Revision 4 to Holtec International Report HI-21 04753 (HI-21 04753)

(Attachment 3 to Reference 6); and Holtec International Report HI-2094416, Revision 1, the NCS analysis for the St. Lucie 2 new fuel vault (Attachment 4 to Reference 6).

As part of Attachment 5 to Reference 1 and in support of the EPU, the licensee submitted Holtec International Report HI-2104753, Revision 1, "St. Lucie Unit 2 Criticality Analysis for EPU and Non-EPU Fuel," which describes the NCS analysis for the St. Lucie 2 SFP storage racks. It describes the methodology and analytical models used in the NCS analysis to show that the storage racks maximum k-effective (keff) will be less than 1.0 when flooded with unborated water for normal conditions, and less than or equal to 0.95 when flooded with borated water for normal and credible accident conditions at a 95-percent probability, 95-percent confidence level.

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-2 The St. Lucie 2 SFP has three fuel storage rack types, with one or more racks for each type:

one Cask Pit Rack (CPR). 6 Region 1 Spent Fuel Racks (SFRs) that contain full-length nonstructural L-shaped stainless steel inserts that serve to enhance local neutron absorption, and 13 Region 2 SFRs. HI-21 04753, Revision 4 proposes 10 storage configurations, 2 in the CPR, 3 in the Region I SFRs, and 5 in the Region 2 SFRs. The CPR contains and credits empty cells, fuel burnup, and the neutron absorbing material BORALTM. The Region 1 SFRs credit fuel burnup. empty cells, control rods, and steel inserts. The Region 2 SFRs credit higher burnups, empty cells, control rods, cooling time, and inserts made from the neutron absorbing material METAMICTM. The presence of soluble boron is credited in the St. Lucie 2 SFP.

New fuel assemblies may be stored in what are normally dry conditions in the St. Lucie 2 new fuel racks in the NFV. It is necessary to update the NFV NCS analysis for the EPU because the maximum planar average enrichment has been increased to 4.6 weight-percent uranium-235.

No credit is taken in fresh or spent fuel storage racks for integral burnable absorbers.

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, Criterion 62 requires, "Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations."

Paragraph 50.68(b)(1) of 10 CFR requires, "Plant procedures shall prohibit the handling and storage at anyone time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water."

Paragraph 50.68(b)(2) of 10 CFR requires, "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. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used."

Paragraph 50.68(b)(3) of 10 CFR requires, "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. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used."

Paragraph 50.68(b)(4) of 10 CFR requires, "If no credit for soluble boron is taken, 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 unborated water. 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."

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-3 Paragraph 50.36(c)(4) of 10 CFR requires, "Design features. 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 this section."

The St. Lucie 2 SFP NCS analysis does take credit for soluble boron for normal operating conditions. Therefore, the regulatory requirement is for the St. Lucie 2 keff 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 keff must remain below 1.0 (subcritical), at a 95-percent probability, 95-percent confidence level, if flooded with unborated water. The St. Lucie 2 NCS uses the double contingency principle to take credit for soluble boron for abnormal/accident operating conditions.

The St. Lucie 2 NFV NCS analysis demonstrated that, consistent with the requirements of 10 CFR 50.68, the maximum keff at full water density was less than 0.95, at a 95-percent probability/95-percent confidence level and that the maximum keff at optimum water density, around 8 percent of full density, was less than 0.98, at a 95-percent probability, 95-percent confidence level.

3.0 TECHNICAL EVALUATION

3.1 Proposed Change There are several proposed TS changes that either impact NCS analyses or implement changes in fuel storage requirements. The reactor operating condition changes related to the EPU affect fuel depletion, which is credited in the spent fuel pool NCS analysis. EPU-related changes impacting NCS analysis may include higher power density, fuel and moderator temperature changes, and soluble boron concentration changes.

One of the TS changes raises the maximum fuel enrichment for fresh fuel storage from a maximum of 4.5 weight-percent uranium-235 to a maximum lattice averaged value of 4.6 weight-percent uranium-235. The TS changes associated with fuel stored in the SFP include increasing the maximum initial enrichment from 4.5 weight-percent uranium-235 to a maximum planar average initial enrichment of 4.6 weight-percent uranium-235, credit for empty storage locations, credit for use of METAMICTM inserts, credit for installation of full-length full-strength, five-fingered control element assemblies and definition of three special configurations referred to in the NCS analysis as inspection and maintenance configurations.

3.2 Method of Review This safety evaluation involves a review of the NCS analyses for the SFP provided as to Reference 6 and for the NFV provided as Attachment 4 to Reference 6 and the related proposed TS changes that were provided as attachments to Reference 3. The SFP NCS analysis includes the effects of the changes in reactor operating conditions associated with the EPU and supports expansion of fuel storage requirements to additional storage configurations. The review was performed consistent with Section 9.1.1 of NUREG-0800.

The staff issued an internal memorandum on August 19, 1998, containing guidance for performing the review of SFP NCS analysis (Reference 2). This memorandum is known colloquially as the 'Kopp Memo' (Reference 9), after the author. While the Kopp memorandum does not specify a methodology, it does provide some guidance on the more salient aspects of OFFICIAL USE ONLY PROPRIETARY INFORM.AJION

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- 4 an NCS analysis, including computer code validation. The guidance is germane to boiling-water reactors and pressurized-water reactors (PWRs), borated and unborated. The Kopp memorandum has been used for virtually every light-water reactor SFP NCS analysis since, including this St. Lucie 2 analysis.

3.3 SFP NCS Analysis Review 3.3.1 SFP NCS Analysis Method There is no generic or standard methodology for performing NCS analyses for fuel storage and handling. The methods used for the NCS analysis for fuel in the St. Lucie 2 SFP are described in HI-2104753, Revision 4. Additional information describing the methods used is provided in the RAI responses attached to References 4, 5, and 6. Some SFP analysis deficiencies were identified during the review, but as will be discussed below, sufficient margin is built into the analysis methodology to offset the deficiencies. Consequently, the methodology is specific to this analysis and, without further revision, is not appropriate for other applications.

3.3.1.1 Computational Methods The LAR seeks to credit fuel assembly burnup in 8 of the 10 defined configurations. The CASMO-4 computer code and its 70-group cross-section library were used to calculate burned fuel compositions and to generate lumped-fission-product cross sections for use with the MCNP-5 computer code. MCNP-5 was used, with its ENDF/B-V and ENDF/B-VI continuous energy cross sections and the CASMO-4-generated lumped fission product cross sections, to calculate keff values. These computer codes and the nuclear data sets with them have been used in many NCS analyses, are industry standards, and are considered acceptable.

Although not common, the computational method used "lumped fission product" (LFP) number densities and cross sections generated by CASMO for use in MCNP. The preparation, testing, and use of LFP are described in Holtec International report HI-2033031 (Reference 7). The use of LFPs was verified by performing equivalent calculations with CASMO-4 and with MCNP-5 using the LFP number densities and cross sections. These calculations were performed to ensure that CASMO-4 cross sections were appropriately converted to the format required by MCNP-5. There is some unquantified and unvalidated uncertainty associated with using the LFP method. The spent fuel analysis includes a "5 percent of the reactivity decrement" uncertainty (Reference 9) to cover lack of validation of spent fuel compositions, including fission products, and a "[ ] of the minor actinide and fission product worth" uncertainty to cover the lack of validation of minor actinides and fission products for calculation of kef!. Recent work published in NUREG/CR-7109 (Reference 8) indicates that an uncertainty of about 3 percent of the minor actinide and fission product worth should be sufficient to conservatively bound biases that may be associated with calculating keff for systems with minor actinides and fission products. The uncertainties adopted for fuel composition and keff calculations are large enough to cover bias and uncertainty associated with use of LFP.

During the review, a question, RAI SRXB-121, was asked concerning how source convergence was checked in the MCNP calculations. According to the RAI response, in all MCNP calculations, results for 50 skipped cycles were used. A review of the Shannon entropy convergence criteria for these calculations showed that in most cases more than 50 skipped cycles would be required to achieve convergence. The RAI response provided an assessment of the impact of revised converged calculations and noted that there was a reduction in the OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 5 margin to the keff limits, with the largest reduction being 0.0052 ~k. This nonconservatism is addressed in Section 3.3.5 of this report.

3.3.1.2 Computer Code Validation Since the NCS analysis credits fuel burnup, it is necessary to consider validation of the computer codes and data used to calculate burned fuel compositions and the computer code and data that utilize the burned fuel compositions to calculate keff for systems with burned fuel.

Consistent with the guidance provided in the Kopp Memo (Reference 9), the analysis has incorporated a "5 percent of the reactivity decrement" uncertainty to cover lack of validation of fuel composition calculations. This uncertainty is calculated as 0.05 times the change in keff from the fresh fuel to the credited final fuel burnup. This uncertainty was calculated by the licensee and applied correctly.

The study used to support validation of keff calculations using MCNP-5 was documented in Holtec International report HI-2104790 (Reference 10). The validation set included 291 critical configurations that included the French Haut Taux de Combustion (HTC) critical experiments (Reference 11). mixed uranium and plutonium critical experiments and some low enrichment uranium fuel pin experiments. During the review it was noted that the validation set included some of the HTC critical experiments that were not recommended for use, that the trending analysis did not evaluate trends in the calculated bias and bias uncertainty associated with variation of plutonium content or with uranium enrichment, and that the wrong value was used for the fresh fuel bias. Thae licensee performed additional analyses to support its response to RAI SRXB-145, showing the impact of excluding some of the HTC critical experiments, and evaluating the variation of the bias with plutonium content and uranium enrichment. This analysis showed that the bias used in the analysis was about 0.0010 too low. Considering the margins available to the regulatory limit, this 0.001 0 ~k nonconservatism is acceptable.

Appropriate critical experiment data was not available to validate keff calculations crediting minor actinides (uranium-236, neptunium-237, and americium-243) or fission products. This deficiency was identified in the analysis. To address this validation deficiency, an uncertainty equal to [ ] of the worth of the minor actinides and fission products was adopted.

Recent work published in NUREG/CR-7109 indicates that an uncertainty of about 3 percent of the minor actinide and fission product worth should be sufficient to conservatively bound biases that may be associated with minor actinides and fission products. Therefore, the [ ]

uncertainty value adequately covers the keff validation deficiencies.

3.3.2 SFP and Fuel Storage Racks 3.3.2.1 SFP Water Temperature NRC guidance provided in the Kopp memorandum states the NCS analysis should be done at the temperature corresponding to the highest reactivity. Analysis was performed with water densities of 0.9787 g/cm 3 and 1.00 g/cm 3 and at water temperatures of 300 OK and 400 OK and at soluble boron concentrations of 0,500, and 1000 PPM by weight. The water densities used reflect the water density range as the water temperature varies from the temperature where maximum water density occurs, near 39 OF, up to 155 OF, which is the assumed maximum normal operating temperature. The 300 OK and 400 OK temperatures were used because the MCNP continuous energy cross sections are available only at a few specific temperatures and MCNP does not do interpolation or temperature corrections. The product of the analysis was a OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 6 storage configuration-dependent temperature bias to be applied to the maximum keff value for that storage configuration.

The NRC staff identified two issues with the SFP water temperature bias determination. The first is that the analysis did not examine water densities between 1.00 and 0.9787 g/cm 3 to ensure that keff did not peak at some value between the maximum and minimum water density values. Instead, engineering judgment was cited by the licensee as the basis for concluding that the calculations were conservative and acceptable. The second issue is that there is currently no way to validate the impact on keff of the variation of cross sections with temperature.

There is some uncertainty associated with utilizing 400 OK cross sections to quantify the impact of cross section variation at 342 OK on the keff value calculated by MCNP.

From Table 7.3 of HI-21 04753, after discarding bias values that would reduce the maximum keff value, the calculated temperature bias values range up to 0.0034 Ak for case 10 with no soluble boron. It is unlikely that the uncertainty on the temperature corrections is as large as 100 percent. Consequently, for purposes of evaluating a balance of conservatisms and nonconservatisms, the SFP water temperature bias determination may be nonconservative by as much as 0.0034 Ak. This nonconservatism is addressed in Section 3.3.5 of this report.

3.3.2.2 SFP Storage Rack Models Three fuel storage rack variations are used in the St. Lucie 2 SFP. All three are manufactured by fabricating boxes that are then welded together at their corners, creating a checkerboard arrangement of storage cells inside fabricated boxes and between boxes. In response to RAI SRXB-130, the licensee stated that the filler panels and corners that are used to complete the periphery of each rack module have the same thickness as the fabricated boxes and are attached such that the minimum cell inner dimension is maintained.

The CPR includes BORAL ' panels held to the exterior side of each box using a steel wrapper.

A detailed representation of the model rack that explicitly models the box, BORAL ' panel, and wrapper was utilized in the analysis. The CPR model used in cases 1 and 8 was a laterally infinite array of fabricated and formed cells. Use of this laterally infinite model, which does not credit neutron leakage from the sides of the CPR rack modules, produces some unquantified margin for the CPR calculations.

The Region 1 and 2 rack modules are also constructed steel boxes that are welded together at the corners, creating formed and fabricated storage cells. However, these rack modules do not include BORAL ' panels and wrappers. A simplified model was used for the Region 1 and 2 rack modules that utilizes a single average repeated cell. This model maintains the center-to center spacing between assemblies and the amount of steel between assemblies. This simplification should have little to no effect on the calculated keff values. The only differences between the Region 1 and 2 rack modules is that an L-shaped steel insert has been placed in all Region 1 cells and that L-shaped METAMICTM inserts may be installed in some Region 2 cells. In both the CPR and the Region 2 racks, with METAMICTM inserts, the minimum boron-10 loading is used in the model.

The staff finds that the design basis models used to model the CPR and Region 1 and 2 racks are appropriate and conservative.

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-7 3.3.2.3 SFP Storage Rack Models Manufacturing Tolerances and Uncertainties The minimum material thicknesses of the box walls and sheathing were used in all design basis calculations. Sensitivity calculations were performed for the CPR rack that determined the maximum center-to-center spacing yielded higher keff values. For the Region 1 and 2 rack modules, the minimum cell inner dimension was used, which is conservative and therefore, acceptable.

3.3.3 Fuel Assembly 3.3.3.1 Bounding Fuel Assembly Design The fuel assemblies used at St. Lucie 2 are all Combustion Engineering (CE) 16x16 or equivalent assemblies manufactured by other suppliers. In some of the older CE 16x16 assemblies, fuel rods were replaced with rods containing B4 C. More recently, the fuel assemblies have included some fuel rods with Gd 2 0 3 mixed in with the U02 in the fuel pellets.

When Gd 20 3 is mixed in the fuel pellets with the U02 , the uranium-235 enrichment is lower than the maximum planar average enrichment for the fuel pellets that do not have Gd 2 0 3 . In both new and old designs, the initial uranium enrichment may be varied from pin-to-pin to control fuel assembly power peaking factors. The fuel assembly design model used in the St. Lucie 2 SFP NCS analysis does not include any B4 C or Gd 2 0 3 rods and utilizes the maximum planar average enrichment for all fuel pins. This increases the amount of uranium-235 present relative to the actual fuel assemblies. Sensitivity calculations were performed to estimate the bias introduced by using the planar average enrichment rather than the actual pin-dependent enrichment variations. The licensee performed sensitivity studies that indicate that modeling the fuel assemblies as having all fuel pellets at the maximum planar average enrichment rather than explicitly modeling the Gd 2 0 3 fuel rods and the replacement of some fuel rods with B4 C rods adds additional unquantified margin to the analysis.

Some fuel assemblies stored at St. Lucie 2 include 6-inch 2.6 weight-percent enriched uranium-235 blankets on the ends of the fuel assemblies. The licensee's calculations were performed both with and without enriched uranium blankets to ensure that the most reactive design was used for each case and, where appropriate, each point on each burnup dependent loading curve. Blankets of different sizes and enrichment were not included in the analysis and therefore are not part of the methodology.

The fuel assembly model included only the active length of the fuel rods. The bounding model did not include fuel assembly grids, nozzles, or the nonfuel ends of the fuel rods. The licensee's sensitivity calculations were performed both with and without soluble boron to show that modeling the grids either reduced reactivity or, when soluble boron was present, caused only a small increase in keff. Not modeling the grids is conservative at unborated conditions, but at some point the amount of soluble boron in the water would make it nonconservative to not model the grids. At 500 ppm of soluble boron, the value used for normal conditions with soluble boron, the impact may be as large as 0.0008 ~k. It is expected that this value will increase as soluble boron concentrations are increased for accident conditions. However, this increase in

~k is offset by the residual uncredited soluble boron.

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-8 3.3.3.2 Fuel Assembly Manufacturing Tolerances and Uncertainties The licensee used an uncommon technique to address fuel storage rack and fuel assembly manufacturing tolerances. In its approach, the licensee used the worst case dimensions for key parameters in the model, effectively incorporating the uncertainty as a bias. It then performed an analysis to show that the margin provided by using the bounding values for a few key parameters was larger than the margin produced by doing detailed uncertainty analysis around the nominal conditions, combining the uncertainty contributions. According to the analysis presented in HI-2104753, including worst case dimensions produces conservative margins varying from 0.0014 to 0.0163 ~k compared to using nominal dimensions and including a more conventional detailed uncertainty analysis. Due to the basis for the method and range of margins observed for the St. Lucie 2 analysis, any future licensing basis changes utilizing this method will require the same comparison of detailed uncertainty analysis results to the simplified method results. For the St. Lucie 2 analysis, the method does generate reasonably conservative margins compared to the more conventional detailed uncertainty analysis.

3.3.3.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on uranium-235 enrichment and various manufacturing tolerances. The manufacturing tolerances are typically manifested as uncertainties, as discussed above, or are bounded by values used in the analysis. These tolerances and bounding values would also carry through to the spent nuclear fuel, common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. The characterization of spent nuclear fuel is more problematic. Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three main areas: a burnup uncertainty, the axial and radial apportionment of the burnup, and the core operation that achieved that burnup.

3.3.3.4 Burnup Uncertainty In the Kopp Memo (Reference 9), the NRC staff provided its method for evaluating burnup uncertainty, "A reactivity uncertainty due to uncertainty in the fuel depletion calculations should be developed and combined with other calculational uncertainties. In the absence of any other determination of the depletion uncertainty, an uncertainty equal to 5 percent of the reactivity decrement to the burnup of interest is an acceptable assumption." The licensee used this approach in HI-2104753 to address the uncertainty in the burned fuel compositions.

3.3.3.5 Axial Apportionment of the Burnup or Axial Burnup Profile Another important aspect of fuel characterization is the selection of the axial burnup profile. At the beginning of life, a PWR fuel assembly will be exposed to a near-cosine axial-shaped flux, which will deplete fuel near the axial center at a greater rate than at the ends. As the reactor continues to operate, the cosine flux shape will flatten because of the fuel depletion and fission-product buildup that occurs near the center. Near the fuel assembly ends, burnup is suppressed due to leakage. If a uniform axial burnup profile is assumed, then the burnup at the ends is over predicted. Analysis discussed in NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis" (Reference 12), has shown that, at assembly burnups above about 10 to 20 GWd/MTU, this results in an underprediction of keff; generally the underprediction becomes larger as burnup increases. This is what is known as the "end effect." Proper selection of the axial burnup profile is necessary to ensure keff is not underpredicted due to the end effect.

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- 9 NUREG/CR-6801 provides recommendations for selecting an appropriate axial burnup profile.

With respect to the axial burnup profile, HI-21 04753 did not use the axial burnup profiles from NUREG/CR-6801. A description of how the axial burnup profiles were derived is provided in Sections 2.3.5.6 and 7.11 of HI-2104753. The NRC staff requested additional information regarding the derivation of the axial burnup profiles, which the licensee provided in its response to RAI SRXB-139 (Reference 6).

The axial burnup profiles used were derived from 744 St. Lucie 2 fuel assembly burnup profiles.

These profiles include 620 pre-EPU profiles and 124 profiles from design calculations for post-EPU cycles. The 744 profiles include 558 profiles for assemblies with enriched uranium blankets and 188 profiles for non blanketed fuel. Separate profiles were prepared for blanketed and unblanketed fuel. Blanketed and unblanketed fuel assemblies are treated differently because application of the relatively lower burnups associated with axial blanket zones to unblanketed assemblies would be significantly and unnecessarily over-conservative.

The process described in HI-21 04753 and the RAI SRXB-139 response results in relative burnup-dependent axial burnup profiles that are more reactive than all of the 744 profiles. One conservative feature of the bounding axial burnup profiles is that they are not renormalized to yield an assembly average relative burnup of 1.0. Instead, use of the profiles conservatively reduces the assembly average burnup by 1.2 to 3.5 percent. The use of these conservative fuel-assembly-design-specific and plant-specific axial burnup profiles is acceptable.

The burnup credit limit curves were derived using polynomial fits that bound the most limiting result obtained using a uniform profile, nonblanket profile and a blanketed fuel profile.

Consequently, separate loading curves for blanketed and nonblanketed fuel are not needed.

3.3.3.6 Planar Burnup Distribution Due to the neutron flux gradients in the reactor core, assemblies can show a radially tilted burnup distribution (Le., differences in burnup between portions or quadrants of the cross section of the assembly). The HI-21 04753 analysis did not consider the effect of planar burnup distribution on reactivity. The impact of radial burnup gradients may be estimated by comparing the distribution of radial burnup tilt information provided in Figure 3-4 of DOE/RW-0496 (Reference 13) with information on the sensitivity of keff to radial burnup tilt provided in Section 6.1.2 of NllREG/CR-6800 (Reference 14). From DOE/RW-0496, the maximum quadrant deviation from assembly average burnup had been observed to be less than 25 percent at low (burnup < 20 GWdlMTU) assembly average burnups and was observed to decrease with burnup, generally being less than 10 percent at burnups above 20 GWd/MTU.

Combining these radial tilt bounding estimates with the keff sensitivity information provided in NllREG/CR-6800, the NRC staff's review of the radial burnup tilts could raise the keff value as much as 0.002 ilk. With the information available, the staff finds that it is conservative to consider this effect as a bias, its potential impact is small, and it is accommodated within the analysis margins.

3.3.3.7 Burnup History/Core Operating Parameters NUREG/CR-6665, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," (Reference 15) provides some discussion on the treatment of depletion analysis parameters that determine how the burnup was achieved. While NUREG/CR-6665 is focused on NCS analysis in storage and transportation casks, the basic prinCipals with respect to the depletion analysis apply generically, since the phenomena occur in the reactor as the fuel is OFFICIAL USE ONLY PROPRIETi\RY INFORMATION

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- 10 being used. The results have some applicability to St. Lucie 2 NCS analyses. The basic strategy is to select parameters that maximize the Doppler broadening/spectral hardening of the neutron field resulting in maximum plutonium-239/241 production. NUREG/CR-6665 discusses six parameters affecting the depletion analysis: fuel temperature, moderator temperature, soluble boron, specific power and operating history, fixed burnable poisons, and integral burnable poisons. While the mechanism for each is different, the effect is similar: Doppler broadening/spectral hardening of the neutron field resulting in increased plutonium-239/241 production. NUREG/CR-6665 provides an estimate of the reactivity worth of these parameters.

The largest effect appears to be due to moderator temperature. NUREG/CR-6665 approximates the moderator temperature effect, in an infinite lattice of high burnup fuel, to be 90 pcmfOK. Thus, a 10°F change in moderator temperature used in the depletion analysis would result in 0.005 ~keff' The effects of each core operating parameter typically have a burnup or time dependency.

For fuel and moderator temperatures, NUREG/CR-6665 recommends using the maximum operating temperatures to maximize plutonium-239/241 production. For moderator temperature, the HI-21 04753 analysis used the post-EPU exit water temperature for the peak power assembly and used a conservatively high fuel temperature. The moderator and fuel temperatures used are therefore acceptable.

For boron concentration, NUREG/CR-6665 recommends using a conservatively high cycle-average boron concentration. The licensee's analysis used a cycle average soluble boron concentration of 750 ppm for pre-EPU cycles and 1000 ppm for post-EPU cycles. The data for S1. Lucie 2 cycles 13 through 18 were used to correctly calculate the average soluble boron concentration for each cycle, yielding a maximum pre-EPU value of 650 ppm, which is well below the 750 ppm value used for pre-EPU fuel burnup calculations. For post-EPU cycles, a value of 1000 ppm was adopted. This value is 300 ppm higher than the predicted boron letdown curve provided in Table F.8 of HI-21 04753 and is to bound future operations. The boron concentrations used are therefore acceptable.

Specific power and operating history are related. Operating history is essentially the history of time spent at various specific power levels. Specific power is a second order effect compared to moderator temperature and soluble boron concentration. The HI-21 04753 analysis used a pre-EPU specific power of 33.5 MW/MTU and a post-EPU specific power of 37.5 MW/MTU. As is described in Section 7.4 of HI-21 04753 and presented in Tables 7.25 through 7.31, sensitivity calculations were performed for +/- 5 MW/MTU. These calculations demonstrated that the St. Lucie 2 SFP rack keff values are only weakly dependent on the speCific power level. The uncertainties associated with specific power and operating history are very small compared to the other uncertainties explicitly included in the analysis and therefore have negligible impact on the overall uncertainty. The specific power and operating history used are therefore acceptable.

3.3.3.8 Integral and Fixed Burnable Absorbers St. Lucie 2 has in the past used fuel assemblies in which some of the fuel rods were replaced with rods filled with B4 C. St. Lucie 2 has also used fuel rods in which some of the rods contained fuel plus Gd 20 3

  • Rather than explicitly modeling the B4C or U02+Gd 20 3 fuel rods, the HI-2104753 analysis used models where all fuel rods were assumed to be at maximum planar average enrichment. As discussed in Section 3.3.3.1 of this evaluation sensitivity studies for St. Lucie 2 SFP indicate this is a conservative modeling simplification.

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- 11 Removable neutron absorbing rods have not been used in the St. Lucie 2 reactor.

3.3.4 Determination of Soluble Boron Requirements Section 50.68 of 10 CFR requires that the keff of the St. Lucie 2 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 unborated water. By considering the double contingency principal, the 1900 ppm of soluble boron that is required by the St. Lucie 2 post-EPU TS can be credited to ensure the St. Lucie 2 keff does not exceed 0.95, provided that the accident and a boron dilution are independent events. Since the minimum required soluble boron has increased from 1720 ppm to 1900 ppm and the soluble boron credited under normal conditions has decreased from 520 to 500 ppm, the pre-EPU soluble boron dilution analysis bounds the post-EPU configuration.

HI-21 04753 considered the following accidents: abnormal temperature, dropped, mislocated, and misloaded fuel assemblies, missing or damaged required METAMICTM insert, missing required control rods, use of an incorrect loading curve, misalignment between the active fuel region and the neutron absorber. The licensee's analysis demonstrated that the fuel storage racks will be subcritical under accident conditions with 1500 ppm of soluble boron, which is below the revised TS required 1900 ppm of soluble boron.

3.3.5 Margin Analysis and Comparison with Remaining Uncertainties This section provides evaluation of additional conservatism in the analysis and evaluation of items that may have been treated nonconservatively.

3.3.5.1 Potential Nonconservatisms The response to RAI SRXB-121 (Reference 6) indicates vthat results from some nonconverged MCNP calculations were used in the analysis. Those new calculations improved convergence and the case-specific uncertainty increased by 0.001 ~k for some cases. This small increase is covered by the conservative margins described below. The nonconverged cases also affected the determination of the loading curves, increasing the keff by up to 0.0052 ~k for some of the points. Since the curves were generated using a target keff + bias + uncertainties of 0.99, the keff of the normal conditions associated with the loading curves may be as high as 0.9952.

Considering the conservative margins described below, the NRC staff finds this acceptable.

As was noted in Section 3.3.3.6, the evaluation of the planar burnup distribution could increase the estimated keff from the St. Lucie 2 NCS analysis by as much as 0.002 ~k.

As was noted in Section 3.3.2.1, the evaluation of the temperature dependence of the calculated keff values did not look at water densities and temperature between the normal condition extremes. It is possible that the calculated keff values may, in some cases, have peaked between the extremes. It is unlikely that the nonconservatism, if it exists, could be as large as 0.0034 ~k. Considering the conservative margins described below, the NRC staff finds this acceptable.

The analysis incorporates a 2.5-percent uncertainty in the fuel assembly average burnup. The response to RAI SRXB-128 (Reference 5) indicates that 1.5 percent of the 2.5 percent covers uncertainty in plant secondary calorimetric power. This leaves 2-percent uncertainty on the assembly relative power. Uncertainty is introduced because only approximately 25 percent of OFFICIAL USE ONLY PROPRIETARY INFORMATION

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- 12 the assemblies are instrumented and that there are uncertainties associated with neutron flux measurement, detector calibration, integration of power over the life of the assembly, and extrapolation of measured data to uninstrumented assemblies. A value more typically used is 5 percent of burnup. Because there is conservatism in the derivation of the axial burnup profiles, and the other conservatisms described in the next subsection, use of the low 2.5-percent uncertainty in the St. Lucie 2 NCS analysis is, therefore, acceptable in the context of the available margins.

3.3.5.2 Potential Analysis Conservatisms The analysis includes several aspects that add margin to the analysis.

These include the following:

  • Axial burnup profiles not renormalized Application of the axial burnup profiles artificially reduces the assembly average burnups by 1.2 to 3.5 percent. This represents additional margin of about 0.003 L\k for an unblanketed assembly burned to 50 GWd/MTU.
  • Burnup coefficient loading curve defined at keff + uncertainties + biases equal to 0.99 Section 50.68 of 10 CFR requires that, when soluble boron in the spent fuel pool is credited, the keff must be below 1.0 when no soluble boron is present. Generating loading curves with a keff of 0.99, without soluble boron, includes 0.01 L\k margin to the analysis.
  • Burnup coefficient loading curve fits either match or conservatively bound the calculated burnup coefficient curve points The final loading curves are second-order polynomial fits that conservatively bound the calculated acceptable loading points. This introduces a small amount of additional margin for most of the range.
  • Conservative treatment of uncertainties An unusual treatment of manufacturing tolerances was used that, according to the analysis, introduced additional margin varying from 0.0014 to 0.0163 L\k.
  • Modeling all fuel rods at the maximum planar average enrichment rather than explicitly modeling B4 C or U02 +Gd2 0 3 fuel rods Many of the fuel assemblies include either fuel rods with U02+Gd 20 3 pellets or with B4 C rods replacing fuel rods. The presence of these Gd rods and B4 C rods is not credited in the analysis. This represents significant unquantified margin. From a generic study presented in Reference 15, not modeling the Gd20 3 represents margin ranging from about 0.08 L\k at zero burnup to 0.002 L\k at high burnups. From Figure 36 in the Reference 15, not modeling the B4 C rods in a CE 14x14 assembly results in margin ranging from around 0.15 L\k at low burnups to 0.002 L\k at burnups around 30 GWdlMTU.

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  • An uncertainty equal to [ ] of the minor actinide and fission product worth was used Recent work documented in Reference 8 indicates that use of an uncertainty as low as 1.S percent of the minor actinide and fission product worth may be appropriate. Use of the [ ] uncertainty adds margin to the uncertainty analysis. The

"[ ] of the worth" uncertainty results in uncertainties ranging from 0.00S4 ilk at low burnups to 0.026 ilk at high burnups. At high burnups, use of the [ ]

uncertainty value increased the total uncertainty by about 0.01 ilk.

3.3.S.3 Conclusion on Analysis of Margins Considering both the potential nonconservatisms identified in Section 3.3.S.1 and the conservatisms identified in Section 3.3.S.2, it is concluded that the available margins offset the potential nonconservatisms.

3.4 The NFV NCS Analysis During the review of the NCS analysis of the SFP it was determined that NCS analysis supporting fresh fuel in the NFV was needed to support related TS changes. Specifically, the TS limiting the fuel stored to a maximum uranium-23S enrichment of 4.S weight percent was changed to permit a maximum lattice average uranium-23S enrichment of 4.6 weight-percent uranium-23S. A NFV NCS analysis, documented in Holtec International report HI-2094416, Revision 1 (Reference 16). was provided as attachment 4 to Reference 6.

This section documents the review of the NFV NCS analysis.

3.4.1 New Fuel Vault (NFV) NCS Analysis Method The analysis is to demonstrate compliance with the following requirements from 10 CFR SO.68:

Paragraph SO.68(b)(2) of 10 CFR requires, "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 0.9S, at a 9S-percent probability, 9S-percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used."

Paragraph SO.68(b)(3) of 10 CFR requires, "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 9S-percent probability. 9S-percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used."

Compliance with these requirements is demonstrated in HI-2094416 by modeling a simplified representation of fresh fuel stored in the NFV submitted by the licensee as part of its application. The licensee's analysis demonstrates that the keff value, including bias and uncertainties, for both full density water and optimum moderation conditions is more than 0.03 ilk below the applicable limits from 10 CFR SO.68.

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- 14 The MCNP4a computer code and nuclear data was used in their analysis to calculate the keff value for fresh fuel in the NFV fuel storage racks. MCNP4a has a long history of use for this type of analysis and is, therefore, acceptable.

MCNP4a and its associated nuclear data were validated in Holtec International report HI-2094486RO, "MCNP Benchmark Calculations," and Section 2 of HI-2094416 states the bias and 95/95 uncertainty from the validation are 0.0013 ilk +/- 0.0086. This bias and bias uncertainty are similar to values reported for other analyses. ConSidering that the values are consistent with similar analyses and that there is more than 0.03 ilk margin to the limits, further review of the validation was deemed unnecessary.

The NRC staff reviewed the computational method and supporting validation and finds them acceptable.

3.4.2 NFV Fuel Storage Racks The steel structures that comprise the NFV fuel storage racks were conservatively not modeled.

Without the steel, the rack model simplifies down to constraints on the spacing and location of the fuel assemblies. All rack structures were modeled as water at the calculation-specific water density.

Section 7 of HI-2094416 states that rack tolerances are not included since the racks are not modeled. Rack tolerances include tolerances on spacing between assemblies and between assemblies and the NFV floor and walls. Not modeling the storage rack steel provides enough margin to cover deficiencies in the uncertainty analysis.

3.4.3 Fuel Assemblies HI-2094416 states that the design basis fuel assembly is a Westinghouse 16x16 and describes the assembly in Table 5.1 of that report. From Table 5.1, the 16x16 assembly is equivalent to a CE 16x16 fuel assembly. The assembly used in the analysis presented in HI-2094416 is consistent with the fuel assembly design used for the spent fuel storage analysis documented in HI-2104753, review of which was covered in Section 3.3 of this safety evaluation.

The fuel assembly design is conservative in that natural uranium blankets and fuel rods containing Gd2 0 3 are modeled as unpoisoned 4.6 weight-percent uranium-235 U0 2. The analysis did not address the impact of using the planar average enrichment simplification. As was shown in Table 7.34 of HI-2104753, the impact of using average enrichments rather than distributed enrichments is small and mixed in direction. Any nonconservatism related to this effect can be covered by the large margins to the limits.

Other than the active lengths of the fuel rods, other fuel assembly components were replaced with water. This is conservative for NVF analysis.

Based on the NRC staffs review of the above, the fuel assembly model used in the NFV NCS analysis is adequately conservative.

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- 15 3.4.4 Analysis of Margins The licensee determined that the maximum keff value, including biases and uncertainties, for the full density flooding case was 0.9152. The applicable limit is 0.95. The margin to this limit is 0.035 .1.k. The licensee determined that the maximum keff value, including biases and uncertainties, for the optimum moderation case was 0.9358. The applicable limit is 0.98. Thus, the margin to this limit is 0.044 .1.k and the NRC staff finds this acceptable ..

A few small potential nonconservatisms were identified in the review. The margins to the limits are large enough to cover potential nonconservatisms associated with deficiencies in the analysis of tolerances and failure to evaluate the impact of some modeling simplifications.

3.5 Conclusions The NRC staff review of the St. Lucie 2 spent fuel storage racks NCS analysis, documented in HI-2104753, and of the St. Lucie 2 NFV NCS analysis documented in HI-2094416, Revision 1, identified some nonconservative items.

Those items were evaluated against the margin to the regulatory limit and what the NRC considers an appropriate amount of margin attributable to conservatisms documented in the analyses.

It was noted that the results from nonconverged Monte Carlo calculations were used in the analysis. The response to RAI SRXB-121 indicated that upon rerunning many of the cases, the reported maximum keff values may be non conservative by as much as 0.005 .1.k. The NRC staff finds this acceptable because the burnup credit loading curves were designed to have a maximum keff' including biases and uncertainties, of 0.99. This left adequate margin to cover the use of results from poorly converged Monte Carlo calculations.

The uncertainty analysis included an allowance of only 2.5 percent of the assembly average burnup used to compare to the burnup credit loading curves. This value is lower than the more typical value used of 5 percent of assembly average burn up.

The effects of keff variation with SFP water temperature and density were quantified only at the extremes of the parameter ranges. Because the analysis did not include intermediate water densities to show that keff did not peak between the extremes of the parameter ranges, the non conservatism associated with this part of the analysis is likely between 0 and 0.0034 .1.k.

The licensee's use of a method for evaluating manufacturing tolerances and uncertainties in its SFP analysis was supported with calculations demonstrating that the revised method yielded conservative total uncertainties with additional margins ranging from 0.0014 to 0.0163 .1.k.

Following review of the supporting analysis reports and based on the margins to regulatory limits, crediting some analysis conservatism, and including consideration of the identified potential nonconservatisms, the NRC staff concludes that there is a reasonable assurance that the St. Lucie 2 SFP and NFV fuel storage racks meet the applicable NCS regulatory requirements.

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- 16 Because of the need to evaluate offsetting effects in the licensee's analysis, its analysis constitutes a methodology of which any change would be a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.

4.0 STATE CONSULTATION

Based upon a letter dated May 2, 2003, from Michael N. Stephens of the Florida Department of Health, Bureau of Radiation Control, to Brenda L. Mozafari, Senior Project Manager, U.S. Nuclear Regulatory Commission, the State of Florida does not desire notification of issuance of license amendments.

5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, 51.33, and 51.35, a draft Environmental Assessment and finding of no significant impact was prepared and published in the Federal Register on January 6,2012 (77 FR 813). The draft Environmental Assessment provided a 30-day opportunity for public comment. The NRC staff received comments that were addressed in the final environmental assessment. The final Environmental Assessment was published in the Federal Register on July 6,2012 (77 FR 40092). Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of this amendment will not have a significant effect on the quality of the human environment.

6.0 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) such activities will be conducted in compliance with the Commission's 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.

7.0 REFERENCES

1. Florida Power and Light Company letter L-2011-021, RL. Anderson, Site Vice President, St. Lucie Plant to USNRC document control desk, re: "St. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, License Amendment Request for Extended Power Uprate, n February 25, 2011 (ADAMS ML110730116).
2. Florida Power and Light Company letter L-2011-466, RL. Anderson, Site Vice President, St. Lucie Plant to USNRC document control desk, re: "St. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Revision to Extended Power Uprate License Amendment Request Proposed Technical Specification 5.6, Design Features - Fuel Storage - Criticality," November 4,2011 (ADAMS ML11314A111).
3. Florida Power and Light Company letter L-2012-181, RL. Anderson, Site Vice President, St. Lucie Plant to USNRC document control desk, re: "St. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Extended Power Uprate License Amendment Request - Supplement to Proposed Technical Specification Changes Related to Spent Fuel Storage Requirements," April 30, 2012 (ADAMS ML12124A065).

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4. Florida Power and Light Company letter L-2011-489, R L. Anderson, Site Vice President, S1. Lucie Plant to USNRC document control desk, re: "S1. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment," December 8, 2011 (ADAMS ML11350A245).
5. Florida Power and Light Company letter L-2012-182, RL. Anderson, Site Vice President, S1. Lucie Plant to USNRC document control desk, re: "S1. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request," May 4,2012 (ADAMS ML12130A478).
6. Florida Power and Light Company letter L-2012-201, RL. Anderson, Site Vice President, S1. Lucie Plant to USNRC document control desk, re: "S1. Lucie Plant Unit 2, Docket No. 50-389, Renewed Facility Operating License No. NPF-16, Response to NRC Reactor Systems Branch Request for Additional Information Regarding Extended Power Uprate License Amendment Request," May 7,2012 (ADAMS ML12132A414).
7. S. Anton, "Lumped Fission Products and Pm148m Cross Sections for MCNP," HI-2033031, Holtec International, January 2011.
8. J.M. Scaglione, D.E. Mueller, J.C. Wagner, and W.J. Marshall, "An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses - Criticality (kef!)

Predictions," NUREG/CR-7109 (ORNLrrM-2011/514), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, April 2012.

9. NRC Memorandum from L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants,"

August 19, 1998 (ADAMS ML003728001).

10. V. Makodym, "Nuclear Group Computer Code Benchmark Calculations," HI-2104790, Holtec International, January 2011 (ADAMS ML11350A245).
11. D.E. Mueller, K.R Elam, and P.B. Fox, "Evaluation of the French Haut Taux de Combustion (HTC) Critical Experiment Data," NUREG/CR-6979, ORNLfTM-2007/083, U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, September 2008.
12. J.C. Wagner, M.D. DeHart, and C.V. Parks, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis," NUREG/CR-6801 (ORNLfTM-2001/273), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2003.
13. C. Kang, et aI., "Horizontal Burnup Gradient Datafile for PWR Assemblies," DOE/RW-0496, U.S. Department of Energy, Office of Civilian Radioactive Waste Management, May 1997.
14. J.C. Wagner and C.E. Sanders, "Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs," NUREG/CR-6800 (ORNLrrM-2002/6), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2003.

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15. CV. Parks, M.D. DeHart, and J.C. Wagner, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel," NUREG/CR-6665 (ORNLlTM-1999/303),

U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, February 2000.

16. B. Brickner, "Criticality Analysis of the St. Lucie Unit 2 New Fuel Vault," HI-2094416, Revision 1, Holtec International, June 2010.

Principal Contributor: Kent Wood Date: September 19, 2012 OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION M. Nazar -2 The NRC has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations, Section 2.390, "Public Inspections, Exemptions, Requests for Withholding." Accordingly, the NRC staff has also prepared a redacted, publicly-available, non-proprietary version of the SE. Copies of the proprietary and non-proprietary versions of the SE are enclosed.

A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA!

Tracy J. Ort, Project Manager Plant Licensing Branch 11-2 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389

Enclosures:

1. Amendment No. 162 to NPF-16
2. Non-Proprietary Safety Evaluation
3. Proprietary Safety Evaluation cc wI Enclosures 1 and 2: Distribution via Listserv Distribution:

PUBLIC RidsNrrDorlDpr LPL2-2 rtf RidsNrrDssStsb RidsNrrDorlLpl2-2 RidsOgcRp RidsNrrLABClayton RidsRgn2MailCenter RidsNrrPMStLucie RidsAcrsAcnw_MaiICTR RidsNrrDssSrxb KWood, NRR ADAMS Accession No. Non-Proprietary SE) ML12263A224 (Proprietary SE ML12243A415 OFFICE LPL2-2/PM LPL2-2/LA SRXB/BC STSB/BC OGC NLO LPL2-2/BC (A) LPL2-2/PM NAME TOrt BClayton CJackson RElliott LSubin JQuichocho TOrt DATE 9118/2012 9/17/2012 9/5/2012 9/12/2012 9/1012012 9/18/2012 911912012 OFFICIAL RECORD COpy OFFICIAL USE ONLY PROPRIETARY INFORMATION

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