ML17089A234

30441R00031, Revision B, MO-99 Target Assembly Nuclear Design for Once-Through Operation.

ML17089A234
Person / Time
Site:	University of Missouri-Columbia
Issue date:	08/26/2016
From:	Choi H General Atomics, Univ of Missouri - Columbia
To:	Office of Nuclear Reactor Regulation, Nordion (Canada), US Dept of Energy, National Nuclear Security Admin, Univ of Missouri - Columbia
References
DE-NA0002773 30441R00031, Rev B
Download: ML17089A234 (134)

Text

ATTACHMENT 6 30441 R00031 Revision B REACTOR-BASED MOLYBDENUM-99 SUPPLY SYSTEM PROJECT M0-99 TARGET ASSEMBLY NUCLEAR DESIGN FOR ONCE-THROUGH OPERATION Prepared by General Atomics for the U.S. Department of Energy/National Nuclear Security Administration and Nordion Canada Inc.

Cooperative Agreement DE-NA0002773 GA Project 30441 WBS 1110

+GENERAL AFOMICS

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B REVISION HISTORY Revision Date Description of Changes A 26AUG 16 Initial Release Revised to incorporate design changes due to redefinition of MURR B

limiting core conditions POINT OF CONTACT INFORMATION PREPARED BY:

Name Position Email Phone Hangbok Choi Engineer Hangbok.Choi@ga.com 858-909-5236 APPROVED BY:

• Name Position Email Phone B. Schleicher Chief Engineer Bob.Schleicher@ga.com 858-455-4733 K. Murray Project Manager Katherine.Murray@ga.com 858-455-3272 K. Partain Quality Engineer Katherine.Partain@ga.com 858-455-3225 DESIGN CONTROL SYSTEM DESCRIPTION R&D DISC QA LEVEL

~

SYS DV&S 181 DESIGN D T&E N II NIA D NA ii

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B TABLE OF CONTENTS REVISION HISTORY .................................................................................................................... 11 POINT OF CONTACT INFORMATION .*...*.................................................................................. 11 DESIGN CONTROL SYSTEM DESCRIPTION ....*...*......*........................... ~ ..................*.......*....* 11 ACRONYMS *...*....................*..........*....***.*.*...*****...***......*.........*...................................................X 1 INTRODUCTION ...........................*.....*.....**..............*..*.....*...................*.......................*.* 1 1.1 Scope of Work ........................................................................................................... 1 1.2 Comparison of RB-MSS with Other System .............................................................. 2 1.3 Physics Design and Analysis Methods ...................................................................... 3 1.3.1 Criticality and neutron flux calculation ............................................................... 3 1.3.2 Depletion calculation ......................................................................................... 3 1.3.3 99Mo production calculation ............................................................................... 3 2 PHYSICS DESIGN REQUIREMENTS .............................................................................. 4 2.1 Design Requirements ................................................................................................ 4 2.1.1 Nominal flux and power distribution of the core ................................................. 4 2.1.2 Reactivity effect due to target rod loading ......................................................... 5 2.1.3 Target rod burnup and management.. ............................................................... 5 2.1.4 Reactivity control system .................................................................................... 5 2.2 MURR Technical Specifications ................................................................................ 5 2.3 Physics Design and Analysis Model... ....................................................................... 7 2.3.1 MURR reference core model ............................................................................. 7 2.3.2 Target assembly loading analysis ................................................................... 12 2.4 Physics Design Computer Codes ............................................................................ 13 2.4.1 MCNP6 model ................................................................................................. 13 2.4.2. MCODE and ORIGEN2 model ........................................................................ 14 2.4.3 Analytic formulation of Mo-99 production ........................................................ 14 3 DETERMINATION OF TARGET LOADING PATTERN ................................................. 17 3.1 Target Rod Model Description ...........................................................:..................... 18 3.1.1 Target assembly dimensions ........................................................................... 19 3.1.2 Material compositions ...................................................................................... 20 3.1.3 Target assembly MCNP6 model.. .................................................................... 21 3.2 Target Assembly Loading Pattern ........................................................................... 24 3.2.1 Selection of bounding values for control blade position .................................. 25 3.2.2 Selection of target rod position ........................................................................ 26 3.2.3 Effect of neutron shield .................................................................................... 27 3.3 Target Assembly at Reference Position .................................................................. 30 4 PHYSICS PERFORMANCE OF TARGET ASSEMBLY*..........................*........**.....*...*. 30 4.1 Target Assembly Criticality ....... :.............................................................................. 32 4.2 Impact of Target Assembly Loading on MURR Core .............................................. 32 4.2.1 Reactivity insertion due to target assembly loading ........................................ 32 4.2.2 MURR core power peaking due to target assembly loading ........................... 33 4.2.3 Core reactivity characteristics .......................................................................... 35 4.2.4 Kinetic parameters ........................................................................................... 36 4.3 Target Assembly Flux and Power Distribution ........................................................ 37 4.3.1 ~t assembly loading ......................................................................... 37 4.3.2 - - target assembly loading ................................................................. 41 iii

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B 4.3.3 Partial Target Assembly Loading ..................................................................... 49 4.4 Sensitivity to Control Blade Position for Base Target Assembly Loading ............... 52 4.4.1 Sensitivity of non-critical core cases ................................................................ 52 4.4.2 Sensitivity of critical core cases ....................................................................... 59 4.4.3 Limiting core configuration ............................................................................... 61 4.4.4 Sensitivity to regulating rod position ................................................................ 62 4.5 Uncertainty Analysis ................................................................................................ 62 4.5.1 Sensitivity calculations ..................................................................................... 63 4.5.2 Total uncertainties of the limiting core performance parameters ..................... 65 4.6 Target Material Depletion ........................................................................................ 66

4. 7 Structural Component Heating ................................................................................ 68

4. 7.1 Steady-state structural component heating ..................................................... 68 4.7.2 Post-shutdown structural component heating ................................................. 70 5 VALIDATION OF METHODS USED IN PHYSICS DESIGN .......................................... 71 5.1 Validation for Computational Methods and Calculations at MURR ......................... 72 5.2 Validation of MURR Control Blade Depletion Model ............................................... 79 5.3 Advanced Test Reactor Criticality Benchmark Problem .......................................... 89 5.4 Verification of Physics Method for Estimating 99 Mo Production ............................... 92 6

SUMMARY

......................................................................................................................94 7 REFERENCES ................................................................................................................ 96 APPENDIX A MURR FUEL ELEMENT PEAKING FACTORS ................................................ A-1 APPENDIX B TARGET ASSEMBLY LINEAR POWER .......................................................... B-1 LIST OF FIGURES Figure 2-1. MURR main chamber ................................................................................................. 7 Figure 2-2. MURR core layout ...................................................................................................... 8 Figure 2-3. MURR map for fuel elements and reflector regions ................................................... 9 Figure 2-4. Typical MURR control blade travel during full power operation ................................ 11 Figure 2-5. MURR control blade travel during 1/13/2014 and 9/15/2015 ................................... 11 Figure 2-6. MCNP6 physics model for the target assembly loading pattern ............................... 12 Figure 3-1. Target assembly schematic ......................................................... '. ............................ 17 Figure 3-2. MCNP6 model of the reference target assembly at axial mid-plane ......................... 22 Figure 3-3. Neutron shield model of the target assembly ........................................................... 22 Figure 3-4. Vertical view of the target assembly ..... :................................................................... 23 Figure 3-5. Horizontal view of the target rod model .................................................................... 23 Figure 3-6. Vertical view of the target rod model ........................................................................ 24 Figure 3-7. Target rod numbers and positions ............................................................................ 24 Figure 3-8. Axial power distribution of the peak power rod for the maximum bum up core ......... 28 Figure 3-9. Azimuthal power distribution of the peak power node for the maximum bumup core28 iv

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Figure 3-10. Axial power distribution of the peak power rod for the extreme burnup core .......... 29 Figure 3-11. Azimuthal power distribution of the peak power node for the extreme burnup core29 Figure 4-1. Target rod I axial neutron flux distribution in position* for the base loading ........ 38 Figure 4-2. Target rod* axial neutron flux distribution in position *for the base loading ...... 38 Figure 4-3. Target assembly azimuthal neutron flux distribution for the base loading ................ 39 Figure 4-4. Power envelope of the base loading for the extreme burnup core .......................... .40 Figure 4-5. Power envelope of the base loading for the maximum burn up core ....................... .40 Figure 4-6. Target assembly pellet linear power distribution ...................... 41 Figure 4-7. Target rod I axial neutron flux for the staggered loading with fresh assembly . . .42 Figure 4-8. Target rod* axial neutron flux for the staggered loading with fresh assembly 1im42 Figure 4-9. Target assembly azimuthal neutron flux for the staggered loading with fresh assembly ............................................................................................................. 43 Figure 4-10. Target rod I axial neutron flux for the staggered loading with fresh assembly 1im43 Figure 4-11. Target rod* axial neutron flux for the staggered loading with fresh assembly 1im44 Figure 4-12. Target assembly azimuthal neutron flux for the staggered loading with fresh assembly i -..........................................................................................................44 Figure 4-13. Power envelope of the staggered loading for the extreme burnup core -

- ......................................................................................................................... 46 Figure 4-14. Power envelope for the maximum burnup core -

-***************************************************************************************************************46 Figure 4-15. Target assembly pellet linear power distribution for the (fresh target-............................................................................................................... 47 Figure 4-16. Power envelope for the extreme burnup core (fresh target

-*********************************** .. *******"****************************************************************"*********47 Figure 4-17. Power envelope for the maximum burnup core (fresh target-............................................................................................................... 48 Figure 4-18. Target assembly pellet linear power distribution (fresh target-...............................................................................................................48 Figure 4-19. Target rod I axial neutron flux for the partial loading .............................................49 Figure 4-20. Target rod* axial neutron flux for the partial loading ........................................... 50 Figure 4-21. Target assembly azimuthal neutron flux for the partial loading .............................. 50 Figure 4-22. Power envelope of the partial loading for the extreme burn up core ....................... 51 Figure 4-23. Power envelope of the partial loading for the maximum burn up core ..................... 51 Figure 4-24. Target assembly pellet linear power distribution for the partial loading .................. 52 v

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Figure 4-25. Variation of target power vs. control blade position ................................................ 54 Figure 4-26. Variation of peak linear power vs. control blade position ....................................... 54 Figure 4-27. Variation of driver fuel peaking factor vs. control blade position ............................. 55 Figure 4-28. MURR control blade age tilt vs age distribution ...................................................... 61 Figure 4-29. Estimation of weekly 99Mo production for different loading patterns ....................... 67 Figure 4-30. Estimation of the target assembly power during - operation ........................ 68 Figure 5-1. A detailed MCNP model of the MURR core ............................................................. 72 Figure 5-2. A depiction of the generalized information flow in the modified MONTEBURNS simulation for predicting hot startup ECPs ................................................................ 75 Figure 5-3. A plot of the predicted ECP vs. actual initial critical height ....................................... 77 Figure 5-4. Cross sectional view of the MURR core ................................................................... 79 Figure 5-5. Axial and radial discretization in the bottom 4" of the MURR MCNP control blade model ......................................................................................................................... 82 10 Figure 5-6. Control blade B reaction rates for; mixed fuel core; 1971 graphite reflector; critical (CB at 17 inches) ....................................................................................................... 84 10 Figure 5-7. Control blade B reaction rates for; mixed fuel core; 2008 graphite reflector; critical (CB at 17 inches) ....................................................................................................... 84 Figure 5-8. kett deviations from critical vs. control blade burnup history for fresh control blades and those blades with depletion history modeled ...................................................... 86 Figure 5-9. Bottom half-inch radial depletion profiles for control blade number 5-08 .................. 87 Figure 5-10. Bottom half-inch radial depletion profiles for control blade number 5-08 ................ 87 10 Figure 5-11. Axial B atom density profiles for different residence times for control blade number 5-08 .............................................................................................................. 88 Figure 5-12. Comparison of MURR and ATR fuel elements ....................................................... 90 Figure 5-13. MCNP model of ATR core~ ..................................................................................... 91 vi

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B LIST OF TABLES Table 2-1. Definition of MURR core burnup .......... :..................................................................... 10 Table 2-2. Nominal process yield of 99Mo production ................................................................. 16 Table 2-3. Nominal process time of 99Mo production .................................................................. 16 Table 3-1. Pellet parameters (cold dimensions) ......................................................................... 19 Table 3-2. Target rod parameters (cold dimensions) .................................................................. 19 Table 3-3. Target assembly parameters ..................................................................................... 20 Table 3-4. Chemical specification of uranium metal ................................................................... 20 Table 3-5. Chemical composition (wt%) of materials for physics model. .................................... 21 Table 3-6. Critical control blade position ..................................................................................... 25 Table 3-7. Expected control blade travelling range for target-loaded core ................................. 25 Table 3-8. Sensitivity to the target rod position (2 fresh target assemblies) ............................... 27 Table 3-9. Reference target rod loading with critical CB position ............................................... 30 Table 4-1. Kett and reactivity insertion values .................................... 33 Table 4-2. MURR core power peaking due to target assembly loading ...................................... 33 Table 4-3. MURR fuel element* power peaking factors************'.********************************************* 34 Table 4-4. Reactivity coefficients of core with two fresh target assemblies ................................ 35 Table 4-5. Reactivity device worth with two fresh target assemblies ........................................... 36 Table 4-6. Kinetic parameters of the core with and without target assemblies ........................... 37 Table 4-7. Calculated target power level and linear power ......................................................... 39 Table 4-8. Calculated target power level and pellet linear power of the staggered loading ........ 45 Table 4-9. Calculated target power level and pellet linear power of the partial loading .............. 49 Table 4-10. Sensitivities of target assembly power and core peaking factors for the base loading53 Table 4-11. Sensitivities of target assembly power and core peaking factors

-***************************************************************************************************************56 Table 4-12. Sensitivities of target assembly power and core peaking factors

-***************************************************************************************************************57 Table 4-13. Sensitivities of target assembly power and core peaking factors

-***********************************************************************************************************************58 Table 4-14. Control blade age model for critical core conditions ................................................ 59 Table 4-15. Sensitivity of target assembly power and peaking factors for the critical core of the base loading .............................................................................................................. 60 Table 4-16. Effect of regulating rod position on target power and peak linear power ................. 62 vii

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 4-18. Uncertainty of core eigenvalue due to simulation and manufacturing ..................... 64 Table 4-19. Uncertainty of target power due to simulation and manufacturing ........................... 64 Table 4-20. Uncertainty of peak linear power due to simulation and manufacturing .................. 65 Table 4-21. Total uncertainty of the performance parameter ...................................................... 66 Table 4-22. Target material composition and bumup for the average bumup core .................... 67 Table 4-23. Component radiation heating of the maximum bumup core .................................... 69 Table 4-24. Spatial distribution of Be reflector radiation heating for the maximum bumup core. 69 Table 4-25. Spatial distribution of Be reflector radiation heating for the minimum bumup core .. 70 Table 5-1. A comparison of predicted initial critical rod height vs. actual at the MURR .............. 76 Table 5-2. A Comparison of the predicted ECP vs recorded rod height data for an actual hot startup from an unscheduled-shutdown event .......................................................... 78 Table 5-3. A Comparison of the predicted ECP vs recorded rod height data for an actual failed hot startup from an unscheduled-shutdown event .................................................... 78 Table 5-4. Deviations for the estimated kert values from critical for the case with fresh control blades (from feasibility study) and with depleted blades ........................................... 85 Table 5-5. Comparison of MURR and ATR core characteristics ................................................ 89 Table 5-6. ATR criticality benchmark calculation ........................................................................ 92 Table 5-7. Comparison of 235 U cross sections ............................................................................ 93 Table A-1. MURR fuel element 1 peaking factor for the maximum bumup core - ) .. A-2 Table A-2. MURR fuel element 2 peaking factor for the maximum bumup core ( - ) .. A-3 Table A-3. MURR fuel element 3 peaking factor for the maximum bumup core - ) .. A-4 Table A-4. MURR fuel element 4 peaking factor for the maximum bumup core ( - ) .. A-5 Table A-5. MURR fuel element 5 peaking factor for the maximum bumup core - ) .. A-6 Table A-6. MURR fuel element 6 peaking factor for the maximum bumup core - ) .. A-7 Table A-7. MURR fuel element 7 peaking factor for the maximum bumup c o r e - ) .. A-8 Table A-8. MURR fuel element 8 peaking factor for the maximum bumup core - ) .. A-9 Table A-9. MURR fuel element 1 peaking factor for the extreme bumup core - ) ... A-1 O Table A-10. MURR fuel element 2 peaking factor for the extreme bumup c o r e - ) . A-11 Table A-11. MURR fuel element 3 peaking factor for the extreme bumup c o r e - ) . A-12 TableA-12. MURR fuel element4 peaking factor for the extreme bumup core-).A-13 Table A-13. MURR fuel element 5 peaking factor for the extreme bumup c o r e - ) . A-14 Table A-14. MURR fuel element 6 peaking factor for the extreme bumup c o r e ( - ) . A-15 Table A-15. MURR fuel element 7 peaking factor for the extreme bumup c o r e - ) . A-16 Table A-16. MURR fuel element 8 peaking factor for the extreme bumup c o r e - ) . A-17 viii

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Table 8-1. Target assembly linear power for the maximum burnup core - ) ............. B-2 Table B-2. Target assembly linear power for the average burnup core - ) ................ B-3 Table B-3. Target assembly linear power for the minimum bumup core ( - ) ......... , .... B-4 Table B-4. Target assembly linear power for the extreme burnup core - }................ B-5 ix

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B ACRONYMS Acronym Description ANL Argonne National Laboratory ANS American Nuclear Society ANSI American National Standards Institute ASME American Society of Mechanical Engineers ATR Advanced Test Reactor Be Beryllium BOC Beginning Of Cycle BOL Beginning Of Life CB Control Blade CHF Critical Heat Flux CHFR Critical Heat Flux Ratio cm Centimeter ECP Estimated Critical Position ENDF Evaluated Nuclear Data File EOC End Of Cycle EOL End Of Life FPO Full Power Day GA General Atomics GTRI Global Threat Reduction Initiative HEU Highly Enriched Uranium HM Heavy Metal HPC High Performance Computing ICSBEP International Criticality Safety Benchmark Evaluation Project ID Inner Diameter INL. Idaho National Laboratory IRPhEP International Reactor Physics Experiment Evaluation Project JENDL Japanese Evaluated Nuclear Data Library kett Effective multiplication factor kW(t) Kilo Watt Thermal LEU Low enriched uranium x

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Acronym Description LOFA Loss Of Flow Accident LP Linear Power LWR Light Water Reactor MCNP Monte Carlo N-Particle MCODE MCNP-ORIGEN Depletion Program MIT Massachusetts Institute of Technology MITR MIT Nuclear Research Reactor Mo02Cl2 Moly Oxy Chloride MURR University of Missouri Research Reactor MWd Mega Watt Day MW(t) Mega Watt Thermal NDR Nuclear Design Report NNSA National Nuclear Security Administration NU REG Nuclear Regulatory Commission Regulation pcm Per Cent Mille ppm Part Per Million RB-MSS Reactor-Based Mo-99 Supply System RMS Root-Mean-Square Tc Technetium U02 Uranium Dioxide U.S. NRC United State Nuclear Regulatory Committee U30s Tri-uranium octoxide wt% Weight Percent Zircaloy Zirconium alloy xi

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 1 INTRODUCTION Technetium-99m (99"1"c) is the most commonly used medical isotope today, accounting for about 50,000 medical imaging procedures daily in the United States [Whipple 2009]. 99"1"c is known to be very accurately penetrating into human body, efficiently detected by scintillation instruments, low residual dose to a patient owing to a short half-life (6.01 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) and flexible in fabricating var_ious tracers depending on diagnostic use. 99Mo is a unique mother isotope of 99 mTc that decays to 99"1"c by 13-emission, becomes 99rc by isomeric transition and stabilizes to 99 Ru. 99 Mo can be produced through neutron capture reaction of Mo. However, the specific activity of the product is very low because of a small cross section of (n,y) reaction. On the other hand, nuclear fission is known to be a more favorable method, which provides a high specific activity and massive production, although it requires radioactive waste treatment of all the fission products and actinides.

Due to the expected shortage of 99Mo supply, National Nuclear Security Administration (NNSA) announced a funding opportunity entitled "Low Enriched Uranium (LEU) fission target technology and accelerator technology for demonstration and full-scale production of a reliable, domestic supply of Molybdenum-99 without the use of Highly Enriched Uranium (HEU)." Based on the projected demand of 99Mo, the technology is required to produce 3,000 6-day curies of 99 Mo per week, steady state, defined as greater or equal to 48 weeks per year.

1.1 Scope of Work The Nuclear Design Report (NOR) describes physics design analysis results of the target rod (or pin)/assembly for 99Mo production. Multiple target rods (i.e., assembly) are used as a component of Reactor-based Mo-99 Supply System (RB-MSS) that produces 99Mo via nuclear reactions. Ttie target rods have been specifically designed to be loaded in the reflector region of University of Missouri Research Reactor (MURR).

The physics analyses have been conducted to determine target assembly configuration, target rod and pellet sizes, pellet enrichment, criticality, power distribution and 99Mo production in conjunction with thermal-mechanical and thermal-hydraulic design analyses. This report describes nuclear performance of target assembly and its impact on MURR core performance.

This report shall be further used for:

• license amendment application to the United State Nuclear Regulatory Committee (U.S.

NRC),

• safety analysis of the target rod/assembly,

• radiation analysis of the target rod/assembly and

• selection of target rod/assembly operational schemes and performance analysis of 99Mo production.

1

ATTACHMENT6 M0~99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 1.2 Comparison of RB-MSS with Other System The target rod uses LEU to produce 99Mo as a fission product of nuclear reaction and is loaded in the reflector region of MURR with its own cooling system. In order to understand design specifications of RB-MSS target rod, a nuclear device which has similar characteristics, in terms of its configuration and constituent materials, is reviewed and summarized here.

The Massachusetts Institute of Technology (MIT) Nuclear Research Reactor (MITR) fission converter is a nuclear fission device of a higher intensity beam, designed to convert neutrons from the MITR to neutrons with a fission spectrum. The fission converter consists of an array of up to 11 MITR fuel elements arranged on a fuel grid plate located in the fission converter tank.

The fission converter design was reviewed by NRC for several design parameters such as criticality, power level and power distribution, which are summarized as follows:

• Subcriticality and self-sustaining chain reaction was evaluated by the Monte Carlo N-Particle (MCNP) code [LANL 2003]. The estimated highest ke!f was 0.670, which is low enough to preclude a criticality accident in the converter and is well below the limit established for MITR fuel storage racks (kett < 0.90). The highest reactivity due to the converter operation is 0.00125 .&k/k, which is within the limit established for the MITR for a movable experiment with a limit of 0.002 8k/k.

• The highest power level of the fission converter is 251 kW(t) or 316 kW(t) depending on placement of the aluminum block when the reactor power level is 10 MW(t). The licensed power level of MITR is 6 MW(t).

• The hot channel factor of the fission converter was estimated by MCNP under the assumption that all power is deposited in the fuel region. The technical specification requires that the nuclear hot channel factor not exceed 1.53 for the fresh fuel condition to satisfy the safety limits and the limiting safety system settings.

The evaluation concluded that the fission converter facility is a complex experimental facility but it is a subcritical array of fuel and cannot maintain a self-supporting chain reaction like a reactor.

In addition, similar to the RB-MSS target assembly, the fission converter cannot perform its irradiation function without the MITR operation. The conclusions were based on acceptance criteria that are stated in several documents, such as NUREG:..1537, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors," and American National Standard ANSl/ANS-15.1-1990, "The Development of Technical Specifications for Research Reactors" (ANS-1 5.1) as applied to experiments [NRC 1999].

2

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 1.3 Physics Design and Analysis Methods The starting point of the physics design is to collect relevant details of engineering design of the target rod, which includes size of the rod, number of target rods, operating conditions, material selection, etc. General Atomics (GA) has conducted independent studies on RB-MSS for various reactor types and target rod/assembly concepts, and narrowed down to a feasible target rod/assembly concept specifically for MURR [Choi 2011, Choi 201 Sa]. The basic method and computer codes used for physics design are outlined.

1.3.1 Criticality and neutron flux calculation The physics calculations are conducted by the Monte Carlo code MCNP6 using Evaluated Nuclear Data File (ENDF)/B-Vll.1[Parsons2012, Conlin 2013] to obtain eigenvalue and neutron flux distribution. The MCNP6 calculations also generate fission energy deposition averaged over a cell (F7 tally) which are used to obtain the driver fuel and target assembly power. Neutron flux averaged over a cell (F4 tally) is used to obtain isotopic cross sections of 235 U, 238U, 239Pu and 99 Mo using tally multiplier for individual segment of the target rod, i.e., pellet. For the physics calculations, 100 million active source histories (10,000x10,000) are used, which gives a standard deviation (1a) of eigenvalue less than 0.01%.

1.3.2 Depletion calculation Depletion calcul~tion of the target assembly is conducted by the MCNP-ORIGEN Depletion Program (MCODE) [Xu 2006], which is an open source linkage-program for combining MCNP6 and the one-group depletion code ORIGEN2 [Croff 1980]. In this coupled simulation, MCNP6 generates neutron flux distribution and cross section tables for the target materials, while the ORIGEN2 code performs the depletion calculation using its full burnup chain and the MCNP6 results, followed by updating target materials consistent with MCNP6 format and repeating this process for multiple depletion steps. It is known that the standalone ORIGEN2 calculation has limited applications because of its cross section library generated for conventional Light Water Reactor (LWR) fuel lattice and burnup [NRC 1997]. The MCNP-ORIGEN coupling enables consistent use of the MCNP cross section data for the depletion calculation with much less computing time when compared to a stand-alone MCNP6 burnup calculation.

99 1.3.3 Mo production calculation 99 The Mo production is calculated analytically by the Bateman equation using microscopic cross sections (capture and fission) obtained from MCNP6. In the analytic equation, the variation of 99 Mo number density is represented by its production and loss rate, where the production includes fission yields from all actinides while the loss includes decay and neutron capture. The detailed equations are provided in Section 2.4.3.

3

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 2 PHYSICS DESIGN REQUIREMENTS The physics design requirements of the target rod/assembly are different from those of commercial or research reactor fuels. However, there are also similarities between these systems such that both systems use nuclear materials to maintain fission reactions and they should satisfy safety limits during normal and transient operations. The system design requirements are driven from system functions which are as follows:

99

• The target rods produce Mo and other desired fission product isotopes at the required rate when subject to neutron irradiation from in the graphite reflector region of the MURR core.

• The target rods maintain the pellet configuration such that the target rod/assembly is safely sub-critical by itself and such that the coupled target rod/assembly and MURR core are safely controllable by the MURR reactivity control system.

2.1 Design Requirements The RB-MSS shall use two graphite reflector positions of the MURR for target rod loading to meet the performance production goal. The maximum allowable 235 U enrichment of the *target material is 19.75 wt%. Before the commercial operation of RB-MSS, prototype target assemblies will be irradiated to demonstrate the 99Mo production process and integrity of the target rod. The ultimate design goals of RB-MSS are:

• The commercial operation shall be capable of delivering 3,000 6-day curies (defined as number of curies 6 days after production) of 99 Mo per week in a steady-state mode, given operation of MURR at 10 MW(t) for

• consecutive hours before shutdown for

• hours per week.

• The prototype operation shall provide 1,500 6-day curies per week.

The physics design and analysis of the target assembly includes evaluation of the target assembly and MURR core performance as summarized below.

2.1.1 Nominal flux and power distribution of the core The nominal power of the MURR core is 10 MW(t) with and without target rod loading. Because the target assembly loading in the reflector region affects the nominal power distribution of the MURR core, the perturbation of the neutron flux and power distribution shall be evaluated and shown to acceptable within the MURR operating limits. The limiting operating conditions of the target assembly in terms of the peak linear power and total power shall be used for the cooling system design and safety analysis of the target assembly system.

4

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 2.1.2 Reactivity effect due to target rod loading The reactivity effects due to target assembly loading as a result of postulated accident conditions shall be estimated, which is necessary to evaluate the performance of existing control/safety system. The reactivity effects that should be known for representative core conditions are the fuel temperature coefficient of reactivity, coolant temperature coefficient of reactivity, coolant density effect (partial and full voiding), sub-criticality of the target rods, and reactivity insertion due to target rod loading. The requirements of the reactivity effect are specified in MURR Technical Specifications 3.1, 3.2, 3.8, and 5.3 which are described in Section 2.2.

2.1.3 Target rod burnup and management The base target operating scheme is llweek - hrs) full power continuous irradiation in two 99 reflector positions. Depending on the Mo demand, the irradiation time could be either shortened or extended. Detailed (pellet-wise) power distribution and burnup of the target rod shall be estimated and used for thermal/mechanical analysis of the target rod.

2.1.4 ~eactivity control system The static reactivity worth of the control rods (blades) must be adequate to permit the power to be quickly adjusted to the desired value even with the target assembly loading. The static reactivity insertion characteristics of the control rods shall be estimated. From the safety view point, the dynamic reactivity and dynamic power transients following a reactor trip in response to reactivity excursion are more important. These transients shall be calculated in the Safety Analysis.

2.2 MURR Technical Specifications MURR Technical Specifications describe the reactivity condition of the reactor and the reactivity worth of control blades and experiments to assure that the reactor can be shut down at all times and to assure that the reactor core safety limits will not be exceeded [MURR 2006]. The limiting conditions for operation, including reactivity limitations, are summarized below.

(3.1.a) The reactor core excess reactivity above reference core condition shall not exceed 0.096 ak/k. Here, Technical Specification 1.27 defines the reference core condition as the condition of the core when it is at ambient temperature (cold) and the reactivity worth of xenon is negligible (< 0.002 Ak/k).

(3.1.b) The reactor shall be subcritical by a margin of at least 0.02 ak/k with the most reactive shim blade and the regulating blade in the fully withdrawn positions.

5

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B (3.2.d) The maximum rate of reactivity insertion for the regulating blade shall not exceed 1.5x10 4 flk/k/sec.

(3.2.e) The maximum rate of reactivity insertion for the four (4) shim blades operating simultaneously shall not exceed 3.0x104 flk/k/sec.

(3.8.a) The absolute value of the reactivity worth of each secured removable experiment shall be limited to 0.006 Wk.

(3.8.b) The absolute value of the reactivity worth of all experiments in the center test hole shall be limited to 0.006 flk/k.

(3.8.c) Each movable experiment or the movable parts of any individual experiment shall have a maximum absolute reactivity worth of 0.001 flk/k.

(3.8.d) The absolute value of the reactivity worth of each unsecured experiment shall be limited to 0.0025 flk/k.

(3.8.e) The absolute value of the reactivity worth of all unsecured experiments which are in the reactor shall be limited to 0.006 flk/k.

(5.3.a) The average reactor core temperature coefficient of reactivity shall be more negative than -6.0x10-5 flk/k/°F (-1.08x104 (Wk)l°C).

(5.3.b) The average core void coefficient of reactivity shall be more negative than -2.ox10..J

~k/k/% void.

(5.3.d) The regulating blade total reactivity worth shall be a maximum of 6.0x10..J flk/k.

Technical Specification 1.35 defines the secured experiments as follows:

"A secured experiment is any experiment which is rigidly held in place by mechanical means with sufficient restraint to withstand any anticipated forces to which the experiment might be subjected to."

Considering the mechanical structure of the target rods and assemblies of RB-MSS and duration of irradiation in the reactor system, the technical demonstration of RB-MSS has been categorized as a "secured experiment". Therefore, Technical Specifications (3.2.e) and (3.8.b) to (3.8.e) are not relevant to the nuclear design analysis of the target rod/assembly.

6

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R0003 1/B 2.3 Physics Design and Analysis Model MURR is a pressurized, reflected , open pool-type, light water moderated and cooled, heterogeneous system designed for operation at a maximum steady-state power level of 10 MW(t). The MURR core consists of eight fuel elements, each having identical physical dimensions. The main chamber of the MURR is shown in Figure 2-1 . Staffs of MURR and GA agreed to use "MURR2015 reflector development MCNP model", dated 2/4/2015, as the reference MCNP model of the MURR core for the target assembly design and safety analysis

[GA 2015].

Figure 2-1 . MURR main chamber 2.3.1 MURR reference core model The MURR2015 model descri bes reactor core and associated facilities that incorporate detailed control blade (CB) geometry and graphite reflectors. The model was established for use in MCNP calculations to obtain the flux profiles and for further use in determining the thermal-hydraulic behavior of the MURR core during upset and accident conditions. In this model, the north-south plane is offset 22.5 degrees clock-wise to facilitate modeling of the individual MURR fuel elements. All annotated descriptions are with respect to this offset axis. Elements of a typical mixed fuel MURR core are considered in this model, i.e., fuel burnup is considered. Also considered are CB age, beryllium (Be) reflector age, reduced water density, and component temperatures. The coolant and pool water densities are - and - g/cm3 ,

respectively. The horizontal and vertical layouts are shown in Figure 2-2 [McKibben 2006].

7

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B The current HEU fuel elements are placed vertically around an annulus between two cylindrical aluminum reactor pressure vessels. The fuel element has 24 curved plates that form a 45-degree arc. The fuel plates are - cm~ inches) thick. The fuel meat is - cm

~inches) thick in each plate and consists of UAlx aluminide fuel containing uranium with a 235 U enrichment of -mo10. The fuel plates are clad with - cm~ inches) of Al-6061 aluminum. The fuel plates are - cm~ inches) long, with an active fuel meat length of

- cm *inches). The plate and meat width varies by plate with each plate having two unfueled edges that are each - cm ~inches) wide.

Water Levol E

M l'i T

E C1>

_...Re f lector Tank 38 ,10mm Figure 2-2. MURR core layout 2.3.1.1 Core configuration The MURR core has a fixed geometry consisting of eight fuel elements, each having identical physical dimensions. The fuel elements are placed vertically around an annulus between two cylindrical aluminum reactor pressure vessels. Cross-sectional views of the MURR reactor core, control blades, reflector, and experimental holes are shown in Figure 2-3.

MURR is currently licensed for a maximum core power of 10 MW(t). This power level provides neutron flux levels in the center flux trap and irradiation positions in the graphite reflector to enable MURR to fulfill its mission of providing experimental and irradiation services to a variety of users. The RB-MSS will install two target assemblies in wedge L (reflector SA) and N (reflector SB).

8

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Figure 2-3. MURR map for fuel elements and reflector regions 2.3.1.2 Driver fuel element loading pattern MURR operates continuously with the exception of a weekly scheduled shutdown. The averaged operation time is approximately. days. hours) per week at full power. A core loading will always consist of four different pairs of elements, with the two elements of each pair loaded opposite of each other in the core. Elements are loaded at beginning-of-cycle (BOC) under xenon-free conditions. Previously irradiated elements are kept in the storage baskets for two or three weeks before being reused in the reactor to allow for decay of the 135Xe. Typically a fuel element will be used in 18 to 20 different core loadings before being retired from the fuel cycle with an average discharge bumup of-150 Megawatt days (MWd).

Fresh elements are always loaded in core positions F1 or FS. In subsequent cycles, an element will usually be alternated between the two positions (i.e., loaded in first in position F1, then in position FS, etc.). The elements will typically be re-loaded in either of these two positions about 4 or 5 times before being transitioned to positions F3/F7 of the core. After 4 or 5 cycles in the F3/F7 positions, the element is alternated between the F2/F6 positions, and then finally discharged from either the F4/F8 position [Stillman 2013].

9

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 2-1 shows the average, minimum, and maximum fuel burnup of the core, which are used for the target design and analysis. The average total core burnup in the equilibrium cycles is roughly 600 MWd. The average burnup of the fuel element is lowest in the F1/F5 positions, followed by F3/F7, F2/F6, and F4/F8. An extreme burnup core concept is also used to represent a core loading pattern that could create the worst power peaking in both the driver and target assemblies. This was a near-maximum burnup core with a fresh and a to-be-discharged element adjacent to each other in the F1 and F8 positions (and F5 and F4), respectively.

Table 2-1. Definition of MURR core burnup Extreme Minimum* Average Maximum Driver fuel burnup core burnup core burnup core burnup core

• <core...:.ext> <core_mill> <core.:._avg> <~ore_max>

Xenon Clean Clean Equilibrium Equilibrium F1 0 0 19 3 F2 117 20 92 122 F3 67 18 60 '

68 F4 142 142 130 145 F5 0 0 19 3 F6 117 20 92 123 F7 67 18 60 68 F8 142 142 130 144 Core total (MWd) 652 360 600 676 2.3.1.3 Control blade position Four CBs and one regulating rod are partially inserted in the core during normal operation. The CB position is affected by the driver fuel depletion, CB absorber material depletion, and degradation of reflector material. Figure 2-4 shows a typical estimated critical position (ECP) of CB (tip distance from fully-inserted CB tip positic;m) during full p.ovver operation. The ECP was generated for Core 14-03 and Core 14-37 in 2014 for the aged and fresh beryllium reflector, respectively. For the typical operation cycle of MURR core, the control blade reaches its equilibrium height about 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (Day-2 state) after startup and slowly withdraws until the end of cycle (EOC).

During 2014 - 2015 operation, the lowest startup and highest shutdown critical position was recorded at CB position of 32.69 cm and 61.6 cm, respectively [Peters 2016]. These two CB positions are used as bounding values that include various reactor perturbations such as Be 10

ATTACHMENT 6 M0-99 Target Assembly N,uclear Design for Once-Through Operation 30441 R00031 /B reflector installation, CB replacement and experimental loading/unloading. Figure 2-5 shows CB movement during 19-month operation in 2014 and 2015. The average CB age during this period is 4.7 to 5.7 years.

Figure 2-4. Typical MURR control blade travel during full power operation Figure 2-5. MURR control blade travel during 111312014 and 911512015 11

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 2.3.2 Target assembly loading analysis The loading pattern of target rod is to place two sets of *rod assembly side-by-side ih the graphite reflector SA and 58 positions of the MURR, shown in Figure 2-6, which are equivalent to Wedge L and N in Figure 2-3. The target assembly loading position was systematically searched such that the peak linear power of the target rod and total target assembly power are within the design limits of the critical heat flux ratio (CHFR) and cooling capability in case of loss of flow accident (LOFA), recommended by the thermal-hydraulic design and safety analysis

[Chiger 2016; Bolin 2016].

Figure 2-6. MCNP6 physics model for the target assembly loading pattern In order to accommodate the neutron flux (power) variations of the target due to MURR operating conditions, the analyses of the target loading pattern are conducted as follows:

• Four core burnup states are used to determine CB bounding positions and to assess the core and target assembly performance.

• The core and target assembly performance is evaluated for both the non-critical and critical core conditions according to the CB insertion.

• For all core states, it is assumed that the target material is fresh without impurities to conservatively estimate the target power from the safety analysis view point.

12

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B

• It is assumed that the Be reflector is fresh to create a higher neutron flux in the target assembly.

• It is assumed that the CB age is tilted such that A & D are fresh and B & C are 8-years I

old to create a higher power peaking in the target assembly.

• It is assumed that the CB tip position is tilted such that the tip position of B & C is higher than that of A & D by 2.54 cm to create higher power peaking in the target assembly.

• The central flux trap is loaded with sample materials to the maximum reactivity during the operation.

• All the experimental holes are plugged with aluminum and silicon for smaller (Inner diameter (ID) < 2.54 cm) and larger holes (ID > 2.54 cm), respectively.

2.4 Physics Design Computer Codes 2.4.1 MCNP6 model MCNP6 is a general-purpose Monte Carlo transport code. A variety of nuclear data libraries have been generated for the continuous energy, discrete, multi-group, thermal and dosimetry*

neutron data [Goorley 2013a; Goorley 2013b]. MCNP6 is used for criticality and neutron flux calculations. In the MURR2015 model, all driver fuel plates are explicitly modeled for fuel meat and cladding of 24x8 fuel plates. The MCNP6 model is briefed as follows:

• A total of 9077 cells and 1362 surfaces are used to construct the full core model.

• Each driver fuel element is modeled by 24 plates with 24 axial numerical meshes. Two axial meshes are grouped into a single material mesh. A total of 2304 materials are defined for the driver fuel.

• The fuel composition have been generated through Argonne National laboratory (ANL)-

MURR MCNP model development program of the MURR HEU fuel cycle [Stillman 2012]. The fuel composition consists of 234U, 235 U, 236U, 238U, 237Np, 238Pu, 239 Pu, 240 Pu, 241 Pu, 242Pu, 241 Am, 1351, 135Xe, 149Pm, and 149Sm and a lumped fission product, which have been obtained by WIMS-ANL [Hanan 1998; Deen 2003; Dionne 2008] and REBUS-3 [Olson 2001] code.

• Each CB (A, B, C and D) is segmented into 292 material zones, resulting a total of 1168 materials.

• Be reflector is modeled by 15 materials (all are fresh).

13

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B

• All beam tubes, flux trap and experimental holes are explicitly modeled.

• The MCNP6 KCODE option is used for the criticality calculation. A total of 100 million active particles are run for each core simulation.

Validation of MCNP code for predicting critical CB position of MURR was conducted for unscheduled shutdown cores after normal operation as described in Sections 6.1 and 6.2. The deviations between the predictions and actual CB heights are less than 1.0% [Peters 2013]. The validation of CB depletion model was also conducted for a 620 MWD MURR core. The error of core criticality was reduced from 1.1 % to 0.41 % when the CB depletion model is applied [Peters 2012b]. Additional benchmark test of MCNP6 summarized in Section 6.3 is a criticality calculation of the Advanced Test Reactor (ATR), of which the fuel type, reflector, and coolant are similar to those of MURR, i.e., HEU aluminide fuel, beryllium reflector, and light water coolant [Kim 2005]. The simulation predicts the criticality of the core within 0.2%Ak.

2.4.2 MCODE and ORIGEN2 model MCODE, as described in Section 1.3.2, is an open source linkage-program for combining MCNP6 and ORIGEN2. The ORIGEN2 code calculates the buildup, decay, and processing of radioactive materials, using the exponential matrix method to solve a large system of coupled linear first-order ordinary differential equations. In this coupled simulation,

• MCNP6 calculates neutron flux distribution and generates cross section tables for the target materials. Each target rod is divided into 50 numerical meshes in the axial direction, while 5 distinct materials are defined vertically for each assembly.

• ORIGEN2 performs the depletion calculation for 1O target materials using its full bumup chain and the cross sections from the MCNP6. The constant neutron flux depletion model is used with time steps automatically selected by the code.

• The target material compositions are updated and fed into MCNP6 for the new flux calculation. The process is repeated for multiple depletion steps.

2.4.3 Analytic formulation of Mo-99 production 99 The Mo production is calculated analytically by Bateman equation [Bateman 1910] using microscopic cross sections (capture and fission) obtained from MCNP6. In the analytic equation, the variation of nuclide number density is represented by its production and loss rate. For example, the 235U, 238 U, and 239Pu number densities can be written as follows:

5 dN- = -o5N5,n

- (2-1) dt 8 "t"

14

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B 8

dN dt

- = -aaNatn a T (2-2)

(2-3) where a!, a!, a;, and a! are 235 U, 238 U, 239 Pu absorption cross sections and 238 U capture cross section, respectively. Then the number densities at time t under neutron flux cp will be as follows:

(2-4)

(2-5)

(2-6) 2.4.3.1 Estimation of 99Mo number density 99 The Mo number density (Nm) in the target pellets during the irradiation period is a balance 99 between Mo production from the fission reactions and the loss due to decay and transmutation as follows:

(2-7) 99 235 238 239 where y5, y8,and y9 are Mo cumulative yields of U (0.06008), U (0.06139), and Pu (0.05982), respectively. The cumulative yield was obtained from Japanese Evaluated Nuclear Data Library (JENDL)-3 instantaneous fission yield [Nakagawa 2003]. Am is a decay constant of 99 Mo (2.92><10-6/sec). er,, rf:, a:. and cf are 235 U, 238 U, 239 Pu fission cross section and 99 Mo absorption cross section, respectively.

After irradiation, target pins are cooled and sent to the collection system for 99Mo extraction.

When extraction starts, the balance of collected 99Mo number density (N°) is written in terms of collection efficiency(~) and decay:

(2-8)

The verification test of the analytical model has been conducted by ORIGEN2.2 code as 99 summarized in Section 6.4, where the prediction error of Mo number density is 0.07% and 0.2% for the target assembly and collection system, respectively [Choi 201 Sb].

15

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 2.4.3.2 Operation scheme of 99 Mo supply system The target pins are continuously irradiated for

• hours per week. The system will shut down 99 when the MURR is shut down for

• hours. The Mo production process includes several steps 99 until the final product ( Mo) is delivered to the end-user in an appropriate physical form . The best estimate of process efficiency (or yield) and process time of each step are summarized in Table 2-3 and Table 2-3, respectively, for the cooling after irradiation , voloxidation for extraction, purification , etc.

99 Volatilization process for separating of Mo from the irradiated uranium is a known process, which converts irradiated uranium pellet into powdered U30 8 [Motojima 1977]. The oxidation process completely changes the physical form from pellets to powder. The crystal structure also changes, which promotes the release of fission products kept within the crystal of uranium 99 dioxide. To improve the efficiency of Mo extraction, a process gas (chlorine) is used, which produces Mo02Cl2. Earlier experiments by Rosenbaum et al. [Rosenbaum 1978] have shown the Molybdenum release rate of 99% by weight without using chlorine gas, when the molybdenum content in the pellet is greater than -300 part per million (ppm). Considering that the process efficiency also depends on the process condition such as temperature and pressure, the collection efficiency of 90% is used in this analysis.

Table 2-2. Nominal process yield of 99Mo production Process Efficiency Collection cell process Extraction cell process Nordion purification process Nordion packaging Overall efficiency Table 2-3. Nominal process time of 99Mo production Process Time (hours)

Cooling after shutdown I Transportation to collection cell and process I Extraction and shipping preparation I Shipping to Nordion I Nordion purification process I Nordion packaging and shipping to customer I Six-day delay to customer Overall reduction factor due to process time *-

16

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 3 DETERMINATION OF TARGET LOADING PATTERN A single target assembly consists of* target rods, which are similar to LWR fuel rods except that the diameter and height of the target rod are much smaller than those of the conventional LWR fuel rod . In order to facilitate target assembly loading/unloading and mechanically stabilize the target rods during irradiation , the thermal/mechanical design implemented several supporting and cooling fixtures around the target rods as shown in Figure 3-1 . From the neutronics design view point, the whole target assembly structure can be divided into:

1) Target rods including U0 2 pellets, cladding, top and bottom end-plugs, and a spacer spring .

2) A cassette that forms a flow channel for target rods, along with top and bottom locking fixture .

3) Aluminum container that provides a coolant flow path.

4) Stainless steel lower water plenum under the target assembly.

5) Neutron shield for power flattening of the target assembly.

Figure 3-1. Target assembly schematic 17

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Current target assembly design has been established through a series of scoping calculations, which include 99 Mo production capability, power distribution and temperature distribution

[Choi 201 Sa]. The design parameters considered during the scoping studies are:

• Target pellet diameter and cladding thickness to estimate the target rod temperature distribution.

• Solid and annular pellet geometry to control peak pellet temperature.

99

• Active target rod height to increase the Mo production rate.

• Variable rod pitches to improve neutron utilization.

• Number of target rods and loading position in the reflector region.

• Reflector material around the target assembly such as graphite, beryllium, and water.

• Axial reflector material of the target rod such as uranium, beryllium and graphite.

• Structural materials such as stainless steel and aluminum to enhance a higher mechanical strength as well as neutron economy.

3.1 Target Rod Model Description Preliminary investigations of the thermal, mechanical, neutronic performance and manufacturing of the target rod recommended that:

• Use a single enrichment and the same dimensions for all target pellets.

• Solid pellets are preferred. The pellet diameter and pellet-cladding gap are kept small with a thin cladding to accommodate high power density.

• Minimize the use of metal around the target rods to promote thermal neutron population.

Aluminum is used as the structural material.

• Use water instead of graphite or beryllium as the moderator.

• Remove top and bottom reflector but keep the spacer spring to minimize the solid waste from the 99Mo collection process.

18

ATTACHMENT 6 M0-99 Target Assembly.Nuclear Design for Once-Through Operation 30441 R00031 /B 3.1.1 Target assembly dimensions Target assembly physical dimensions and materials are summarized in Table 3-1, Table 3-2 and Table 3-3 for the pellet, target rod and target assembly, respectively. The target pellet has dishes and chamfers to reduce the mechanical interaction between the pellet and cladding.

Because the pellet is modeled as a solid cylinder in the physics analysis, effective isotopic number densities are used for the pellet composition. It should also be noted that the target rod full height is slightly different from actual height ~ cm) in the numerical model.

Table 3-1. Pellet parameters (cold dimensions)

Pellet material Pellet density (%)

Pellet 235U enrichment range (wt%)

Pellet diameter (cm)

Pellet height (cm)

Dish height (cm)

Shoulder width (cm)

Chamfer height (cm) ---

Chamfer width Pellet surface roughness Target rod full height (cm)

Target rod active height (cm)

Table 3-2. Target rod parameters (cold dimensions)

Cladding material Cladding thickness (cm)

Cladding surface roughness Gap fill material Gap width (cm)

Zircaloy-4 (Zirc-4)

End-plug material *Zirc-4 19

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00031/B Table 3-3. Target assembly parameters Target rod Number of rods Number of coolant holes Coolant hole pitch (cm)

Coolant hole diameter, maximum (cm)

Coolant channel Coolant material Coolant density (g/cm 3 Coolant flow rate (m/s)

Cassette material

)

Cassette thickness, minimum (cm)

-* water Supporting fixture Structural material Top flange upper section (gram)

Top flange middle section (gram) ---

Aluminum 6061-T6 3.1.2 Top flange lower section (gram)

Cassette base plate (gram)

Material compositions Table 3-4 and Table 3-5 summarize material compositions of uranium and structural materials (Zircaloy-4, Aluminum 6061-TS and stainless steel 316L) used for physics modeling. The uranium data is from Y-12 standard specification of LEU [Parker 2015]. In the physics analysis, impurities are not included to conservatively estimate the target power. The Zircaloy-4 specifications are provided by a manufacturer [ATI 2016]. The aluminum and stainless .steel data are from the American Society of Mechanical Engineers (ASME) specifications

[ASME 2011].

Table 3-4. Chemical specification of uranium metal Element Uranium purity U-232 Units weight%

µg/gU ---LEU U-234 U-235 U-236 Total impurities Equivalent boron content (EBC) weight%

weight%

µg/gU

µg/gU ppm 20

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00031/B Table 3-5. Chemical composition (wt%) of materials for physics model Zircaloy-4 Aluminum Stainless Steel Material type R60804 A96061 316L Density (g/cm3) 6.56 2.7 8.03 Aluminum 96.68 Carbon 0.03 Chromium 0.1 0.195 17.0 Copper 0.275 Iron 0.21 0.7 66.395 Magnesium 1.0 Manganese 0.15 2.0 Molybdenum 2.5 Nickel 12.0 Oxygen 0.125 Phosphorus 0.045 Silicon 0.6 1.0 Sulfur 0.03 Tin 1.45 Titanium 0.15 Zinc 0.25 Zirconium 98.115 Total 100 100 100 3.1.3 Target assembly MCNP6 model The MCNP6 model of the reference *rod target assembly is shown in Figure 3-2 at axial mid-plane followed by a schematic of the neutron shield as shown in Figure 3-3 along with the target rod position. The horizontal view includes

• target rods, coolant channel, cartridge, aluminum housing and pool water. The vertical configuration of the target assembly and housing is shown in Figure 3-4. The vertical view doesn't show all the target rods, but it shows overall flow path through the aluminum housing and stainless steel low plenum. Figure 3-5 and Figure 3-6 show the horizontal and vertical views of target rod, respectively. The target pellet, cladding, pellet-cladding gap and upper plenum are explicitly modeled. The target rod numbering is shown in Figure 3-7, where target assemblies 1 and 2 reside in the MURR reflector SA and SB positions, respectively.

The target assemblies have been embedded in the MURR core model as follows:

• 1000 cells are used to define target pellets. Two pellets are defined as a single cell. The fuel gap, dish and chamfer voids are smeared into pellet.

21

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B

• 100 surfaces are used to define the pellet, cladding, coolant and other structure.

• 10 materials are used for target pellets. 10 materials are used to define other structure.

Figure 3-2. MCNP6 model of the reference target assembly at axial mid-plane Figure 3-3. Neutron shield model of the target assembly 22

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Figµre 3-4. Vertical view of the target assembly Figure 3-5. Horizontal view of the target rod model 23

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Figure 3-6. Vertical view of the target rod model Figure 3-7. Target rod numbers and positions 3.2 Target Assembly Loading Pattern The target assemblies are planned to be installed in reflector - positions as shown in Figure 2-6. Depending on MURR operating condition, especially the CB insertion depth, the flux level in the reflector region changes. In order to comply with the neutron flux variations and to maintain the cooling capability of the target assembly, the power density of the target assembly should be carefully determined by examining appropriate target rod positions (distance) from the 24

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/8 core. The target power is typically peaked at the axial middle zone. In order to reduce the axial power peaking and to increase the overall target power, a neutron shield is attached to the cartridge.

3.2.1 Selection of bounding values for control blade position The MURR core has a one-week operation cycle~ During each cycle, the CB travels in and out of core to compensate for excess reactivity of the fresh fuel and reactivity loss due to depletion.

The actual CB insertion depth depends on the driver fuel burnup, CB absorber material depletion, reflector material degradation and any reactivity perturbations. Typically, the CB travels over 1O cm during first two days after startup, and then very slowly withdraws as fission products saturate. The driver fuel also has two distinct states: clean fuel and equilibrium xenon.

The critical CB positions of the equilibrium core are summarized in Table 3-6 for the core with and without target assemblies. When the target assemblies are loaded, CBs are inserted more deeply into the core. The estimated additional CB insertion depth is 2 to 10 cm depending on the core states. When these offset values are applied to the lowest and highest CB position during 2014-2015 operation, the estimated lowest and highest critical CB position with a 2-target assembly loading are 33.7 and 51.9 cm, respectively (given in Table 3-7). The CB offset values of <core_min> and <core_avg> were applied to BOC and Day-2 cases, respectively, while that of <core_max> was used for EOC CB position. Here, <ci>re_max> doesn't necessarily mean the EOC state, but it at least considers burnup advancement of the core. In the actual simulation of the core with target assemblies, the lowest CB position was further extended to 44, 40, 30 and 30 cm for the maximum, average, minimum and extreme bumup core, respectively.

Table 3-6. Critical control blade position Extreme .. * . < *'M.inimum .Average***

• Maximum butnupcore burnupeor~

,. .. .. . burnup cpre ' l:)urnup core .

,i*

- *. ~J' <core_ref> <cori!-':.min> *. .<:c::ore_avg> <core.;;;.rilax~.. -

Xenon state clean clean equilibrium equilibrium Regulating rod (cm) 25.4 25.4 38.1 38.1 Critical CB position without 41.67 35.09 59.87 65.0 target assembly Critical CB position with 2 fresh 38.75 32.75 52.89 55.28 target assemblies CB offset (cm) 2.92 2.34 6.98 9.72 Table 3-7. Expected control blade travelling range for target-loaded core Recorded critical CB position*

without tar et loading

• 25 Expected critical c;:e r~llge with

• -:target loading . - **

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B

.Lowest Highest Lowest Highest BOC 36.6 42.6 33.7 39.6 Day-2 49.1 58.8 42.1 51.8 EOC 53.6 61.6 43.9 51.9 3.2.2 Selection of target rod position The performance of the target assembly (i.e., 99 Mo production) depends on the target power which is determined by the target uranium enrichment and neutron flux level (or position) in the reflector. Table 3-8 shows a comparison of target performance for two different rod positions:

. The simulations have been conducted with lowest CB position for 4 core burnup states to consider power peaking in the low half of the core. The 99 Mo production was estimated in terms of 6-day Curies per week for 2 target assemblies which are irradiated for -

• It should be noted that the 99 Mo production is only for sensitivity purposes, because the production rate is seriously over-estimated here due to CB age and mechanical tilts.

The recommended upper limit of the peak linear power is ~ kW/m from the thermal-hydraulic design. Considering uncertainties associated with neutronics design (see Section 4.6), it has been recommended to keep the peak linear power at ~ kW/m. For the 4 core bumup states with the lowest CB position, the lowest peak linear power of target rods at - is - kW/m, while the highest peak linear power of target rods at - is - kW/m. Therefore, an intermediate distance has been chosen as the reference position for the target assembly loading, i.e.,

26

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 3-8. Sensitivity to the target rod position (2 fresh target assemblies)

Rod position Target power Peak linear Mo-99 production Core burnup state (cm) (kW) power (kW/m) (6-day Ci/week)

Maximum with equilibrium xenon Average with equilibrium xenon ** --- ---

Minimum without xenon Extreme without xenon **

3.2.3 Effect of neutron shield The effect of neutron shield is shown in Figure 3-8 and Figure 3-9 for the axial and azimuthal target linear power distributions, respectively, of the maximum burnup core with 2 fresh target assemblies. The axial power is plotted for the center rod ), while the azimuthal power is plotted through the axial mid-plane ). The neutron shield reduced the peak linear power of the target from* t o

• kW/m - % ).

It should be noted that the peak power rod position is different from each other for the non-peak linear power node (i.e., rod number ** axial node *>

shielded and shielded assemblies, and so is the peak linear power position (axial node). For the of the non-shielded assembly, the local linear power is reduced by*% at that specific node when the target assembly is shielded.

Figure 3-10 and Figure 3-11 show axial and azimuthal target power distribution of the extreme burnup core. The axial peak power position is down-shifted due to CB and regulating rod deeply inserted into the core, but the power is still effectively controlled by the neutron shield.

27

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B l

-+- no shield

...... shielded j

0 5 10 15 20 25 Axial node Figure 3-8. Axial power distribution of the peak power rod for the maximum burnup core Figure 3-9. Azimuthal power distribution of the peak power node for the maximum bum up core 28

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B I...

Cll

~

...c.

ftl Cll c

-+- no shield

- - shielded 10 15 20 25 Axial node Figure 3-10. Axial power distribution of the peak power rod for the extreme bumup core Figure 3-11. Azimuthal power distribution of the peak power node for the extreme bumup core 29

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 3.3 Target Assembly at Reference Position The key performance parameters of the target assembly are summarized in Table 3-9 for target assembly loading at the reference position (i.e., - cm from the core center). The -

99 Mo production is given for two loading patterns: base and staggered. In case of the base loading, two fresh target assemblies - ) are loaded and discharged - - For the staggered loading, one target assembly - ) is loaded and discharge - in an alternative way after - irradiation in each loading position.

Table 3-9. Reference target rod loading with critical CB position Mo-99 production (6-day

- Target power Peak linear Ci/week)

Core bumup state (kW) power (kW/m) Staggered Base loading loading Maximum with equilibrium xenon Average with eauilibrium xenon Minimum without xenon Extreme without xenon 4 PHYSICS PERFORMANCE OF TARGET ASSEMBLY Definitions of Target Assembly Loading Pattern and Operation The reference target assembly consists of

• target rods, aluminum cartridge and stainless steel shield. In addition to the base and staggered loading mentioned in Sec. 3.3, the target assembly can also be partially loaded to comply with the demand. The target assembly loading patterns and corresponding operating scenarios are summarized as follows:

• Base loading deploys I fresh target assemblies at the reference loading position and discharges - assemblies - -

• Staggered loading deploys 1 fresh target assembly in one reflector for -

irradiation. After-* another fresh target assembly is deployed in the other reflector and irradiated for - - Both assemblies are at the reference loading position.

• Partial loading uses a reduced number of target rods (i.e., < rods per assembly) and the operation scheme is the same as that of the base loading. Target 30

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B rods are symmetrically loaded in from the center rod position, i.e.,

(see Fig. 3-7).

Target Power and 99/Vlo Production 99 The Mo production is calculated from the target power. For C!ll the physics simulations, the target power is calculated as follows:

• The target pellet is fresh

• without impurities except for the staggered loading which has one irradiated target assembly. The initial coi:nposition of the irradiated target assembly doesn't include impurities either.

• The target assembly power is obtained by normalizing MURR core power to 10 MW thermal. Specifically, F7 (fission energy deposition) tally of the MCNP6 is used to edit the MURR core and target assembly power.

• The 99Mo production calculation assumes that the neutron flux is constant throughout the irradiation. This will results in an over-prediction of 99 Mo production. The 99Mo production is estimated for the average bumup core which is the most probable core during normal operation. Unlike core performance analysis, the 99Mo production calculations are conducted with aged Be reflector and no CB age or mechanical tilt.

• For the staggered loading, though the target power is calculated based on one fresh and one irradiated target assembly, the 99Mo production is approximately calculated from the I I base loading case by irradiating assemblies for weeks and taking I of the total production.

Evaluation of Target Assembly Performance and Core Characteristics The performance of the target assembly and the associated MURR core characteristics have been evaluated for the key physics parameters as follows:

• criticality of the stand-alone target assembly;

• core coupling impact between the target assembly and MURR core (reactivity and power);

• core coefficients of reactivity and reactivity device worth;

• neutron flux and power of the target assembly for different operation schemes;

• effect of control blade movement on the target assembly performance;

• uncertainties due to numerical simulation and manufacturing;

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 99

• target material depletion and Mo production.

4.1 Target Assembly Criticality Conservative calculations of kett for two target assemblies were performed using. the MURR core model by replacing the MURR driver fuel and cladding with water and assuming the control blades and regulating rod are completely withdrawn from the core. The calculated value of kett for the target assemblies respectively. When both - are loaded, the kett is indicating that the target assemblies will be subcritical with a large margin for uncertainty. The uncertainties are given in 95% confidence level, i.e., 2 standard deviations (2a).

4.2 Impact of Target Assembly Loading on MURR Core Loading two target assemblies in the MURR reflector region causes perturbations in neutronics and thermal performance of the MURR core as follows:

• The MURR core excess reactivity will increase, which will be compensated by the reactivity control devices

• The neutron flux and power of the - fuel elements will increase even thougtJ the total MURR driver fuel power is maintained at 10 MW.

• The thermal power generated by the target assemblies won't affect the MURR core cooling capability. The heat flow from the target assembly to the MURR pool will be negligible (i.e., s 20 kW).

4.2.1 Reactivity insertion due to target assembly loading The reactivity insertion due to target assembly loading is summarized in Table 4-1. The reactivity worth was calculated by replacing the water in the target assembly cartridge with fresh, cold target rods. Under the hot operating condition, the reactivity insertion due to a single assembly loading is less than 0.31% .6.k/k. The maximum reactivity insertion due to hot target rod is the same as that of the cold target rod for all core burnup states.

32

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 4-1. Kett and reactivity insertion values with Target i n . Target i n .

-fully Core burnup state assembly loaded kett Llklk (%) kett Llklk (%)

housing Maximum with 0.99995 0.99368 0.99658 0.29 0.99660 0.29 equilibrium xenon Average with 0.99999 0.99392 0.99673 0.28 0.99677 0.29 equilibrium xenon Minimum without 0.99995 0.99385 0.99662 0.28 0.99690 0.31 xenon Extreme without 1.00002 0.99382 0.99667 0.29 0.99685 0.31 xenon 4.2.2 MURR core power peaking due to target assembly loading The target assembly loading in the MURR reflector has a relatively small impact on power peaking of the MURR core driver fuel element. Table 4-2 compares the distribution of MURR fuel element power with and without target assembly loading. For most of core states, the fuel element peaking factor is the highest because the fuel burnup is the lowest and the fuel is located close to the target assembly.

Table 4-2. MURR core power peaking due to target assembly loading Number of Inner-most fuel plate Outer-most fuel plate Core burnup target state Peaking Peaking rods Axial node* Axial node factor factor Maximum with I equilibrium xenon I

Average with equilibrium xenon Minimum without

• I xenon Extreme without xenon

• Axial nodes -

• I correspond to the bottom and top of MURR drive fuel plate, equally spaced.

For the

• fuel element, the peaking factor of the inner plate is always higher than that of the outer plate. When* target assemblies are loaded, the increase of peaking factor. is less than 33

ATTACHMENT 6 .

M0-99 Target Assembly Nuclear Design for Once-Through Oper~tion 30441 R00031 /B 4% for the inner plate, while the maximum increase of peaking factor is - for the outer plate. However, the absolute value of the outer plate peaking factor is always less than that of the inner plate peaking factor. More detailed distributions of peaking factors for* element are given in Table 4-3, where the peaking factors are defined as follows:

• Element peaking = Fuel element power I (total core power/8)

• Radial peaking = Fuel plate average power density I Fuel element average power density

• Axial peaking =Axial node power density I Fuel plate average power density

• Total peaking =Element peaking x Radial peaking x Axial peaking Table 4-3. MURR fuel element* power peaking factors Minimum burnup Av~rage burnup Max.imum burnup Extrem~ burnup core core* core. core 1.189 1.177

• 1.105 1.129 Radial Axial Total Radial Axial Total Radial Axial Total Radial Axial Total I 1.859 1.346 2.975 1.867 1.379 3.030 1.798 1.264 2.511 1.837 1.271 2.636 I 1.508 1.356 2.431 1.516 1.374 2.452 1.473 1.269 2.066 1.486 1.269 2.129 I 1.301 1.343 2.077 1.308 1.375 2.117 1.270 1.275 1.789 1.276 1.266 1.824 I 1.163 1.351 1.868 1.169 1.384 1.904 1.139 1.275 1.605 1.135 1.271 1.629 I 1.068 1.341 1.703 1.072 1.371 1.730 1.044 1.281 1.478 1.041 1.272 1.495 I 1.000 1.360 1.617 1.007 1.383 1.639 0.979 1.274 1.378 0.974 1.269 1.395 I 0.950 1.345 1.519 0.959 1.384 1.562 0.930 1.270 1.305 0.925 1.267 1.323 I 0.915 1.348 1.467 0.921 1.380 1.496 0.895 1.268 1.254 0.890 1.265 1.271 I

0.887 1.344 1.417 0.892 1.381 1.450 0.868 1.262 1.210 0.862 1.271 1.237 0.867 1.349 1.391 0.871 1.387 1.422 0.850 1.264 1.187 0.844 1.265 1.205 0.851 1.348 1.364 0.857 1.389 1.401 0.834 1.266 1.167 0.832 1.264 1.187 0.838 1.344 1.339 0.845 1.391 1.383 0.825 1.268 1.156 0.822 1.264 1.173

• • 0.831 0.827 1.348 1.358 1.332 1.335 0.839 0.832 1.385 1.389 1.368 1.360 0.818 0.818 1.273 1.275 1.151 1.152 0.815 0.812 1.262 1.161 1.257 1.152

• • 0.827 0.830 1.345 1.348 1.323 1.330 0.830 0.833 1.388 1.406 1.356 1.379 0.818 0.823 1.276 1.281 1.153 1.165 0.813 0.819 1.259 1.156 1.269 1.173

• • 0.837 0.851 1.359 1.357 1.352 1.373 0.840 0.852 1.414 1.417 1.398 1.421 0.833 0.851 1.278 1.285 1.176 1.208 0.830 0.847 1.266 1.186 1.274 1.218

• • 0.876 0.909 1.365 1.374 1.422 1.485 0.872 0.907 1.423 1.435 1.460 1.532 0.879 0.922 1.277 1.288 1.240 1.312 0.876 0.919 1.275 1.261 1.283 1.331

• • 0.963 1.044 1.377 1.396 1.577 1.733 0.955 1.035 1.454 1.475 1.634 1.797 0.985 1.079 1.304 1.309 1.419 1.561 0.981 1.078 1.296 1.435 1.298 1.580

• • 1.174 1.396 1.413 1.444 1.972 2.397 1.159 1.371 1.501 1.532 2.048 2.472 1.224 1.472 1.315 1.325 1.779 2.155 1.231 1.491 1.304 1.812 1.317 2.217 34

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00031/B 4.2.3 Core reactivity characteristics 4.2.3.1 Coefficient of reactivity Though the MURR core power distribution, or power peaking factor, is altered when two target assemblies are loaded, the core reactivity characteristics won't change due to target assembly loading. The reactivity coefficients of the core with two fresh target assemblies were calculated for the fuel temperature coefficient, coolant temperature coefficient and void reactivity as summarized in Table 4-4.

The MURR fuel temperature coefficient was calculated by increasing the fuel temperature by 906.4°C, i.e., from 20.44°C to 926.84°C. The variation of fuel temperature coefficients is consistent with core burnup state and kept negative. Due to the limitation in cross section data, however, the temperature dependence of lumped fission products (all fission products except for 135 1, 135Xe, 149Pm, 149Sm) was not considered.

The coolant temperature coefficient was calculated by changing the coolant temperature from 20.44°C to 76.84°C, where the c0olant density changes from 0.99815 to 0.97378 g/cm3 . The principal cross section data used for the calculations are basically the same for both the lower and higher coolant temperature conditions, but the S(a,13) thermal scattering was correctly treated. The coolant void reactivity of the core was calculated by removing 99.9% of coolant from the core.

The statistical uncertainty (2a) of the reactivity coefficient is -2.6x104 when the error propagation rule is used. However, the perturbations of the fuel temperature and coolant density were large enough to obtain consistent results. The least negative values are -1.03x10"6 and -

1.69x104 /1k/k/°C for the fuel and coolant temperatur~ coefficient, respectively.

Table 4-4. Reactivity coefficients of core with two fresh target assemblies Fuel temperature Coolant void Coolant temperature Core burnup state coefficient < reactivity coefficient (Ak/k/°C)

{Ak/k/°C) CAk/k/o/ovoidina)

Maximum with

-1.20x10.s -1.69x104 -3.40x10"3 eQuilibrium xenon Average with

-1.16x10.s -1.73x104 -3.40x10-3 equilibrium xenon Minimum without

-1.27x10-6 -1.85x104 -3.40x10"3 xenon Extreme without

-1.03x10.s -1.79x104 -3.40x10"3 xenon 35

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 4.2.3.2 Regulating rod worth The impact of target assembly loading on the existing reactivity devices (CB and regulating rod) were estimated for their reactivity worth and subcriticality margin as summarized in Table 4-5.

Reactivity worth of regulating rod was calculated from fully-in and fully-out conditions while the CB is kept at its critical position. For the selected core states, the lowest worth of the regulating rod is 3.41 x10"'3 Ak/k.

4.2.3.3 Control blade subcriticality margin The CB subcriticality margin was at first calculated for the average bumup core. Among four CB's (A, B, C and 0), the subcriticality margin of C is the largest even though it is almost the same as that of B. The subcriticality margin increases as the core burnup increases due to effective neutron flux changes. The lowest subcriticality margin with most reactivity CB (A) and regulating rod fully withdrawn is 0.055 Ak/k for the minimum burnup core.

Table 4-5. Reactivity device worth with two fresh target assemblies Subriticality margin of Regulatfog rod worth Core burnup state control blade (L\k/k) (L\k/k)

Maximum with equilibrium xenon 0.109 3.55x10"'3 Average with equilibrium xenon 0.106 3.41x10-3 Minimum without xenon 0.055 3.69x10-3 Extreme without xenon 0.076 4.02x10-3 4.2.3.4 Core excess reactivity The excess reactivity of the core with two fresh target assemblies was calculated for the minimum burnup core which has the largest excess reactivity. All CB's and regulating rod are fully withdrawn from the core. The excess reactivity of the cold core (all in room temperature) is 0.072 Ak/k, while that of the hot operating core is 0.067 Ak/k.

4.2.4 Kinetic parameters The kinetic property of the core is dominated by the driver fuels, which is slightly perturbed by the presence of two target assemblies in the reflector region . The effective (or adjoint weighted) neutron generation time (/\eff) and effective delayed neutron fraction (J3eff) can be readily obtained from the MCNP6 calculation [Kiedrowski 2010]. The results are summarized in Table 4-6 for different burnup states of the core with and without target assemblies. The neutron generation time tends to increases when the two target assemblies are loaded, but the difference is very small. The effective delayed neutron fractions of the core with and without 36

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B target assemblies coincide within the uncertainty range (+/-20). From the viewpoint of point kinetics, the target assembly loading won't deteriorate the slope of power increase (or inverse reactivity period) during the reactivity-induced transient.

Table 4-6. Kinetic parameters of the core with and without target assemblies Target Corebumup loading state Aett (IJsec) STD (1a) Pett STD (1a)

Maximum 62.3 0.273 0.00723 0.00015 Average 60.6 0.278 0.00731 0.00015 No target Minimum 53.7 0.244 0.00749 0.00016 Extreme 57.0 0.248 0.00732 0.00015 Maximum 62.7 0.237 0.00730 0.00013 Average 61.5 0.235 0.00745 0.00014 Two target assemblies Minimum 54.7 0.215 0.00766 0.00014 Extreme 58.4 0.220 0.00769 0.00014 4.3 Target Assembly Flux and Power Distribution 4.3.1 Base target assembly loading The neutron flux in the target assembly is driven by incoming neutrons from the core even though the target assembly produces neutrons from fission reactions. Figures 4-1 and 4-2 show 4-group axial neutron flux distributions of the target rods 6 and 17 (see Fig. 3-7 for rod numbering) when two fresh target assemblies are loaded. The upper energy boundaries of the 4-group are 20 MeV (Group 1), 0.1 MeV (Group 2), 5.53 keV (Group 3), and 0.625 eV (Group 4). The solid and dotted lines indicate the MURR core states <core_ext> and

<core_max>, respectively. The axial nodes - correspond to the bottom and top, respectively. It is obvious that the neutron flux drops at the bottom and top section of the target rod. The neutron flux is suppressed even more in the top section due to the control blades, which is relaxed to a certain extent when the control blades are pulled out in the maximum burnup core. The thermal neutron flux is effectively flattened in the vertical middle section, from node -

• owing to the neutron shield.

Figure 4-3 shows neutron flux in the azimuthal direction, from rod - (position * ) and from rod - (position .). at axial middle plane. The neutron population is suppressed at the assembly edge region due to elongated neutron absorber on the cartridge edge. For the 37

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B average burnup core , the peak neutron fluxes of the target pellet are

1. /cm 2 *s for group 1, 2, 3 and 4, respectively.

r~ .**

• Group 1

• Group 2 Ii *

• a a

• Group 3 '

a a 0 I * '

• a a

• Group 4

• a a "

• 0

• a a *

• a

-+- D- - - - 1 a core_max>

• a a

. ..l -+

~

0

..... *l*it:. *

• t

• I t f t I * .-.- .. fl t t I-5 10 15 20 25 Axial node Figure 4-1. Target rod I axial neutron flux distribution in position

• for the base loading

• Group 1

• Group 2

~.

., a a I a

a

• ...._ a a Group 3 Group 4 iii *j a a a C'i

• 0 a

• a

~

E 0

<core~ma x

• a

)(

a a

i

r.

• a z

c

...0

:i GI 0

0

• t ** ** ** * *

• 0 .  :

0 t

• 1 a

I I 0 0 5 10 15 20 25 Axial node Figure 4-2. Target rod

• axial neutron flux distribution in position

• for the base loading 38

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Figure 4-3. Target assembly azimuthal neutron flux distribution for the base loading Calculated target assembly power and pellet linear power are given in Table 4-7. Figures 4-4 and Figure 4-5 show the pellet linear power envelope of the extreme and maximum burnup core, respectively. The target axial power profile is bottom-peaked for the extreme burnup core and changes to middle-peaked shape for the maximum burnup core as the regu lating rod and control blades are withdrawn from the core. The distribution of pellet linear power is shown in Figure 4-6 for the four different MURR core states. For the most probable operating core condition , i.e., the average burnup core, the peak pellet linear power and total target assembly power are Table 4-7. Calculated target power level and linear power Extreme Minimum Average Maximum burnup core burnup core burnup core 1----------------------t~----. ----t---- ~--~-- ----+---

Peak linear power (kW/m)

Rod number Axial node'*

'*Axial nodes - correspond to bottom and top node, respectively. All nodes are equally spaced.

39

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B T-0 5 10 15 20 25 Axial node Figure 4-4. Power envelope of the base loading for the extreme burnup core

~

~

QI

~

0

...c..

"'c:

GI I

J t

0 5 10 1-15 20 25 Axial node Figure 4-5. Power envelope of the base loading for the maximum burnup core 40

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R0003 1/B

• <core_ max>

• <core_avg>

<core_min>

<core_ext>

• -- 1111- - - - 1 Pellet linear power (kW/m)

+

Figure 4-6. Target assembly pellet linear power distribution 4.3.2 - target assembly loading Prior to the commercial operation of target assemblies (full loading), a prototype operation of 99 target assembly will be conducted to demonstrate the Mo production capability and performance of the target rod . The prototype target rod and assembly are exactly the same as those for commercial operation . However, the prototype operation is supposed to adopt different loading patterns such as staggered or partial loading.

irradiation under normal operating condition. After - operation, another target assembly is loaded in the other reflector position.

Figure 4-7 and Figure 4-8 show 4-group axial neutron flux distributions of the target rods I

-

• respectively, when a fresh target assembly is loaded in position lllt the extreme and maximum bumup core . The axial flux shape of the target assembly is almost the same as that of the base loading. Figure 4-9 shows the azimuthal flux variation on the axial middle plane, which shows a slight decrease of neutron flux level in reflector * .

41

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Ui' N

~

E

<core_ext>

T 0

0 0

0 0 0

0 "

II Ii 1* _. "

-t 0 0 0

0 Group Group Group Group 1

2 3

4

.s

• 0 core_max>

0

" 0

)(

I * *

~ "

• 0

..e c

• ** ** ** ; *** *

• ii ... ...

I QI z .t.

• * ._ . -++ .- ' I I *' ;

• I 0 0  :

0 0

-i t- i i J ' :

5 10 15 20 25 Axial node Figure 4- 7. Target rod I axial neutron flux for the staggered loading with fresh assembly . .

• Group 1

• • • " ~

• Group 2

• Group 3 0 0

• 0 Group 4 0 0

• * +

0 0 <core_max>

0 I

• 0 0

0

• 0 0

• a I *

~ .....,..~

.. I 0

0 i

• I

......

• t ~

I

• • i

• 0 0

0 0 5 10 Axial node 15 20

• 25 Figure 4-8. Target rod . axial neutron flux for the staggered loading with fresh assembly . .

42

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Figure 4-9. Target assembly azimuthal neutron flux for the staggered loading with fresh assembly . .

For the staggered loading with a fresh target assembly . . (a --burned assembly is I

• the 4-group axial neutron flux distributions of the target rods - are shown in Figure 4-10 and Figure 4-11 , respectively. The azimuthal fluxes are shown in Figure 4-12.

• Group 1

• * * *

• Group 2 D

o ---+

a

• Group 3

• Group 4 0

• f. a D

• D a

D

. a a

t A 0

0 0 0 5 10 15 20 25 Axial node Figure 4-10. Target rod I axial neutron flux for the staggered loading with fresh assembly . .

43

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B T

• Group 1 I

• Group 2

<core_ext>

0 0

0 D

<core_max>

• I *

• Aa a a

• a o

0

• Group 3 D

Group 4

• 0 0

D D D -I t l t t *

i * * ****

0 5 10 15 Axial node Figure 4-11. Target rod . axial neutron flux for the staggered loading with fresh assembly . .

Figure 4-12. Target assembly azimuthal neutron flux for the staggered loading with fresh assembly ~

Target assembly power and rod linear power of the staggered target assembly loading are summarized in Table 4-8 when the fresh assembly is in reflector

• The linear power envelopes are shown in Figure 4-13 and Figure 4-14 for the extreme and maximum burnup core , respectively. The distribution of pellet linear power is shown in Figure 4-15. The axial power profile is the same as that of -

-

• but the overall linear power is slightly reduced in the burned assembly. For the most 44

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B probable reactor condition, i.e., average burnup core, the power of the target assembly . . is lower by

• kW when compared with that of the target assembly . .. which is due to the target fuel depletion. The total target assembly power - is slightly lower than that of the base target assembly loading by 1%.

The pellet linear power envelopes of the staggered loading with fresh target . . are shown in Figures 4-16 and Figure 4-17 for the extreme and maximum burnup core, respectively, and their distribution is plotted in Figure 4-18. The total target assembly power The power from the target assembly . . is higher by

• kW when compared with that from the target assembly . ..

Table 4-8. Calculated target power level and pellet linear power of the staggered loading Core burnup Target power state Maximum with equilibrium xenon Peak linear power (kW/m)

Rod number Axial node Assembl Average with equilibrium xenon Peak linear power (kW/m)

Rod number Axial node Assembl Minimum without xenon Peak linear power (kW/m)

Rod number Axial node Extreme without xenon Peak linear power (kW/m)

Rod number Axial node 45

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B

-r---

0 5 10 15 20 25 Axial node Figure 4-13. Power envelope of the staggered loading for the extreme burnup core -

e

~

~

~

CD

~

0

...ca Q.

CD c

25 Axial node Figure 4-14. Power envelope ***I for the maximum burnup core -

46

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B T

• <core_ max>

• <core_avg >

• <core_min >

• <core_ext>

_J

....0 Q)

.c E

J z

Pellet linear power (kW/m)

Figure 4-15. Target assembly pellet linear power distribution for the (fresh target-I .

E'

~

~

...GI

~

0 Q.

"'c GI 0 5 10 15 20 25 Axial node Figure 4-16. Power envelope for the extreme burnup core (fresh target 47

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 0 5 10 15 20 25 Axial node Figure 4-17. Power envelope * * * * *

• for the maximum burnup core (fresh target

<core_avg>

<core_ min>

<core_ext>

. .L.

• ..i *

. .. . .. .a.....

..... . + *

• mli* -

• Pellet linear power (kW/m)

Figure 4-18. Target assembly pellet linear power distribution (fresh target -

48

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 4.3.3 Partial Target Assembly Loading The neutron flux and power distribution of the target assembly were simulated The rod positions - and - were selected to maintain the symmetry. Other rod positions are loaded with solid stainless steel 316L rods (filler rods) to avoid unnecessary neutron thermalization while maintaining the nominal flow condition .

The 4-group axial neutron flux distributions of the target rods - are shown in Figure 4-19 and Figure 4-20, respectively. The azimuthal fluxes are shown in Figure 4-21. Target assembly power and pellet linear power are summarized in Table 4-9. The pellet linear power envelopes of the extreme and maximum burnup core are shown in Figure 4-22 and Figure 4-23, and their distribution is plotted in Figures 4-24.

Table 4-9. Calculated target power level and pellet linear power of the partial loading Extreme Minimum Average Maximum burnup core burnup core bumup core burnup core t--~~~~~~~~~~-r-~- -~--~- -~-+-~- -~-+-~-

W)

Peak linear power (kW/m)

Rod number Axial node T

Group 1 Group 2

• a* a

• *

• a a a

g *

• a a

Group 3 Group 4 Vi

• l'i E

a a a

• i* a a

~)(

• a a

s c

0

._,i

.... 6i

s z

Cl>

0 *

--+!-  !

• *** a a a ; ;

0 5 10 15 20 25 Axial node Figure 4-19. Target rod I axial neutron flux for the partial loading 49

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00031/B

• Group 1

• Group 2

• ** 0 0 0

0 Group Group 3

4

• 0 0

• • a a

• 0

-A~,._.,,____~___t t .. *

~ 8 i 8 s ***

0 5 10 15 20 25 Axial node Figure 4-20. Target rod

• axial neutron flux for the partial loading Figure 4-21. Target assembly azimuthal neutron flux for the partial loading 50

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B I

l...

Ill QI c

J 0 5 10 15 20 25 Axial node Figure 4-22. Power envelope of the partial loading for the extreme bumup core

~

- t - - - - - + - _ ___,,.____ i A ' '

• a 0 5 10 15 20 25 Axial node Figure 4-23. Power envelope of the partial loading for the maximum bum up core 51

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B

• <core_ rnax>

• <core_avg>

• <core_rnin> +

<core_ext>

Pellet linear power (kW/m)

Figure 4-24. Target assembly pellet linear power distribution for the partial loading 4.4 Sensitivity to Control Blade Position for Base Target Assembly Loading The peak linear power of the target rod is dominated by the CB insertion depth . The sensitivity of peak linear power to the CB position was calculated for the maximum, average , minimum and extreme core bumup states. The CB position represents the core response to all kinds of reactivity perturbations to the core such as the fuel depletion and material aging. In order to determine limiting cases for the thermal-hydraulic and cooling system design, the effects of CB position on the peak linear power and total target assembly power have been estimated for both the non-critical and critical cores.

4.4.1 Sensitivity of non-critical core cases The sensitivity calculations have been conducted for a wide range of CB position beyond the estimated minimum and maximum critical CB position for the base target assembly loading pattern (described in Section 3 .2 .1 ). In this simulation, the regulating rod was fixed to its typical position: 25.4 cm for the extreme and minimum burnup core and 38 .1 cm for the average and maximum burnup core. It should be noted that the CB movement is synchronized with the regulating rod in real operation . Therefore, the actual operating range of CB will be smaller than the values used for simulations.

The results are summarized in Table 4-10 for the total target power, peak target linear power (LP) and driver fuel peaking factors. The total target power is relatively low for the BOC condition which is a relatively short term period. For the equilibrium core (average and maximum burnup cores), the target power stays within . kW. The peak linear power of the target rod is kept between - kW/m for the simulated equilibrium core states and CB 52

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B position. The variations of total target power and peak linear power versus CB position are shown in Figure 4-25 and Figure 4-26, respectively.

Table 4-10. Sensitivities of target assembly power and core peaking factors for the base loading CB position Target Peak LP Inner plate Outer plate Core state (cm) power (kW) (kW/m) peaking peaking 62 2.573 2.147 60 2.577 2.173 58 2.605 2.192 Maximum 56 2.622 2.208 burnup with 54 2.239 equilibrium 52 2.674 2.273 xenon 50 2.709 2.302 48 2.755 2.348 46 2.792 2.385 44 2.861 2.441 56 2.488 2.109 54 2.505 2.131 52 2.529 2.163 Average 50 2.575 2.192 burnup with 48 2.609 2.240 equilibrium xenon 46 2.648 2.263 44 2.684 2.310 42 2.757 2.374 40 2.802 2.392 42 2.790 2.386 40 2.857 2.419 Minimum 38 2.898 2.453 bumup without 36 2.944 2.489 xenon 34 32 3.055 2.576 30 3.120 2.609 42 2.906 2.301 40 2.930 2.351 Extreme 38 2.991 2.403 bumup without 36 3.053 2.475 xenon 34 32 30 .. - 3.130 3.182 3.231 2.539 2.594 2.636 53

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B l

r l

Equilibrium xenon case x x x No xenon case j 0

a o a  : o

.j

-4 30 35 40 45 50 55 60 65 CB position (cm)

Figure 4-25. Variation of target power vs. control blade position l

~

- - Equilibrium xenon case -

l...

0 x *

• x x

0 I

1 a 0 a 6

' 8 i 0

a 0

Cl) c

)(

JC

)(

No xenon case 35 40 45 50 55 60 65 CB position (cm)

Figure 4-26. Variation of peak linear power vs. control blade position The MURR fuel element power peaking occurs in the fuel element** The fuel element peaking is higher for the BOC state when the CB and regulating rod are deep into the core. The calculated maximum peaking factor is 3.231 for the inner plate. For the outer plate that faces the target assembly, the maximum peaking factor is 2.636, which is lower than that of the inner plate for all core state and CB positions. The variations of fuel element peaking factor versus CB position is shown in Figure 4-27.

54

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B The maximum linear power found from the sensitivity calculation is

• kW/m for the extreme burnup core with CB positioned at 30 cm and regulating rod positioned at 25.4 cm , which is recommended for the thermal-hydraulic, critical heat flux (CHF) margin analysis. The maximum target assembly power is - kW for the maximum burnup core with CB at

• cm and regulating rod at

• cm, which is recommended for the target assembly cooling analysis.

I Inner plate

...0 (No xenon case)

Inner plate ti Ill (Equilibrium xenon case )

• *a a*

~

Cl II c

ii:

a 0

• a* *

• Ill Cll I ** ** ** * ** ** *

• Q.

Cll t !I Q.

"i

~

...=

Cll c

35 40 45 50 55 60 65 CB position (cm)

Figure 4-27. Variation of driver fuel peaking factor vs. control blade position The sensitivities of non-critical core to the CB position have also been calculated -

. Tables 4-11 performance parameters

- *Table 4-13 summarizes the can be seen that the total target assembly power and peak linear power of these cores are always lower than those of the base loading case when the CB is at its estimated lowest position.

55

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 4-11. Sensitivities of t a r l i i i i i i i i l i i i i l d core peaking factors -

CB position Target Peal< LP Inner plate Outer *plate Core state (cm) power(kW) (kW/m) peaking peaking 62 2.571 2.140 60 2.-588 2.159 Maximum burnup with 58 56 54 --- --- 2.603 2.643 2.668 2.184 2.198 2.227 equilibrium 52 2.678 2.265 xenon 50 2.723 2.306 48 2.777 2.322 46 44 56 --- --- 2.811 2.863 2.490 2.376 2.410 2.098 Average 54 52 50 --- --- 2.502 2.535 2.582 2.128 2.148 2.187 burnup with 48 2.612 2.223 equilibrium xenon 46 2.664 2.264 44 2.690 2.306 42 40 42 --- --- 2.769 2.810 2.789 2.342 2.382 2.373 Minimum burnup without 40 38 36 --- --- 2.825 2.906 2.959 2.416 2.453 2.492 xenon 34 3.009 2.529 32 3.072 2.551 30 3.120 2.601 Extreme 42 40 38 --- --- 2.901 2.952 3.014 2.282 2.351 2.396 bumup without xenon 36 34 32 30 3.047 3.129 3.193 3.238 2.456 2.522 2.577 2.637 56

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 4-12. Sensitivities of t a r i i i i i i i i i l i i ' d core peaking factors -

CB position Target Peak LP Inner plate Outer plate Core state (cm) power (kW) (kW/m) peaking peaking 62 2.555 2.144 60 2.568 2.163 Maximum burnup with 58 56 54 --- --- 2.602 2.611 2.663 2.180 2.209 2.239 equilibrium xenon 52 50 48 --- --- 2.677 2.719 2.761 2.267 2.294 2.332 46 44 56 --- --- 2.802 2.853 2.477 2.364 2.417 2.109 Average 54 52 50 --- --- 2.497 2.541 2.582 2.136 2.163 2.190 burnup with 48 2.603 2.232 equilibrium xenon 46 2.673 2.266 44 2.701 2.295 42 40 42

..- --- 2.750 2.780 2.789 2.357 2.396 2.384 Minimum burnup without 40 38 36

......- --- 2.844 2.911 2.939 2.419 2.457 2.490 xenon 34 3.023 2.544 32 30 42 40 3.066 3.123 2.895 2.935 2.578 2.614 2.284 2.340 Extreme bumup without xenon 38 36 34 32 30 3.005 3.076 3.122 3.165 3.244 2.402 2.476 2.519 2.583 2.630 57 J

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00031/B Table 4-13. Sensitivities of target assemb-r and core peaking factors CB position Target Peak LP Inner plate Outer plate Core state (cm) power (kW) (kW/m) peaking peaking 62 2.555 2.070 60 2.566 2.073 Maximum burnup with 58 56 54

--- --- 2.583 2.619 2.646 2.100 2.117 2.141 equilibrium xenon 52 50 48

-..-- --- 2.662 2.702 2.739 2.177 2.193 2.236 46 44 56

..- --- 2.792 2.852 2.472 2.284 2.338 2.029 Average 54 52 50

--- --- 2.509 2.525 2.552 2.038 2.080 2.093 burnup with 48 2.606 2.138 equilibrium xenon 46 2.620 2.177 44 2.683 2.205 42 40 42

..-- --- 2.731 2.772 2.778 2.255 2.295 2.327 Minimum burnup without 40 38 36

--- --- 2.830 2.891 2.933 2.363 2.407 2.442 xenon 34 3.000 2.468 32 30 42 40 3.040 3.125 2.897 2.927 2.511 2.542 2.200 2.266 Extreme burnup without xenon 38 36 34 ..-- ---

3.005 3.046 3.119 2.332 2.379 2.454 32 30

- 3.158 3.222 2.501 2.550 58

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 4.4.2 Sensitivity of critical core cases During normal operation, the reactor is critical all the time. However, it is difficult to model a critical core without knowing the status.of driver fuel burnup, CB depletion, Be reflector age, etc.

Therefore, a simplified model is proposed in this analysis to simulate critical cores with different CB positions: the CB age has been changed from 0-year to 8-year. Table 4-14 shows the CB age matrix used for the critical core calculations. The critical core was searched for ke!f value of 1+/-0.00006.

For all the critical core cases of the base target assembly loading shown in Table 4-15, the calculated total target power and the maximum linear power are less than those of non-critical core cases. For the extreme burnup core, the peak linear power is - and - kW/m for the non-critical and critical core, respectively~ when the average CB age is the same (4-year) for both cases. The peak linear power is the highest for the CB average age of 4-years which has the largest age tilt in the CB age model (Table 4-14). If the CB age is 8-years, the peak linear power drops to - kW/m even though the critical CB position is deep into the core, i.e.,

32.28 cm, because the CB age tilt is assumed to be zero. Therefore it is crucial to use the appropriate CB age model when estimating target assembly power to avoid unnecessary over-design of the target assembly.

The investigation of MURR operation history of the CB age has shown that the average CB age during the operation period of April 2006 and October 2016 is 3.65-year. During this period, the maximum CB age tilt between AD and BC is 6.3-years, while the maximum tilt of any CB pair is 9.2-years. The analysis also has shown that the CB age tilt is the highest when the CB average age is 4- to 5-years and the tilt is small when the CB average age is very low (-1-year) or high

(-7-years) as shown in Figure 4-28. In summary, the average CB age used for the simulation is close to the actual operating data. The CB age tilt used for the simulation is higher than the operation data, but could be conservatively used. The non-critical core analysis that uses the CB age tilt of 8-years will be the conservative estimation of the target power.

Table 4-14. Control blade age model for critical core con ditions CB C'!ntrol blade age (year)

A 0 0 0 4 8 B 0 4 8 8 8 c 0 4 8 8 8 D 0 0 0 4 8 Average 0 2 4 6 8 59

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 4-15. Sensitivity of target assembly power and peaking factors for the critical core of the base loading CB Target Inner Outer Core CB age Peak LP position kett power plate plate state (year) (kW/m)

(cm) / *(kW} peaking peaking 0 58.86 1.00000 2.622 2.208 Maximum burnup 2 57.61 1.00002 2.617 2.213 with 4* 55.28 0.99995 2.637 2.218 equilibrium 6 52.41 1.00002 2.642 2.220 xenon 8 47.71 1.00005 2.678 2.227 0 56.74 0.99998 2.537 2.134 Average burnup 2 55.25 0.99994 2.535 2.147 with 4* 52.89 0.99999 2.511 2.155 equilibrium 6 50.08 1.00000 2.530 2.155 xenon 8 45.40 0.99995 2.534 2.154 0 38.06 1.00004 3.039 2.511 Minimum 2 36.12 1.00005 3.032 2.523 burn up 4* 32.75 0.99995 3.031 2.565 without xenon 6 30.36 1.00003 3.027 2.536 8 26.69 1.00005 3.065 2.497 0 43.65 1.00003 3.004 2.366 Extreme 2 41.88 1.00001 2.984 2.389 burn up 4* 38.75 1.00002 2.976 2.397 without xenon 6 36.26 1.00005 2.991 2.400 8 32.28 0.99995 3.008 2.376 60

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Figure 4-28. MURR control blade age tilt vs age distribution 4.4.3 Limiting core configuration The critical core calculations have shown that the non-critical ce>re calculations conservatively estimate the target power and the maximum linear power. Hqwever, it should also be noted that the critical core simulation here is limited by the range of CB age and, therefore, the variation of CB position is smaller when compared with the expected CB traveling range (see Section 3.2.1 ). This means that consideration of CB age is not sufficient to realistically model the critical core. It is also true that in the actual core the CB position could be even higher or lower than those used for the critical core calculations due to Be reflector aging and other reactivity perturbations. And it is also logical to assume that the target power and its axial shape are dominated by the CB position. Therefore, the limiting core configuration has been selected from the non-critical core cases as follows:

• For the total target power, the maximum burnup core with the equilibrium xenon and CB at* cm is used. The corresponding total target power is . . kW.

• For the peak linear power, the extreme burnup core without xenon and CB at

• cm is used. The corresponding peak linear power is - kW/m.

The power peaking factor of the driver fuel and linear power of the target rod are given in Appendices A and B, respectively, for the limiting core configurations.

61

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 4.4.4 Sensitivity to regulating rod position The effect of regulating rod position was assessed for critical cores. Three regulating rod positions, i.e., lowest (25.4 cm), highest (38.1 cm) and middle (31.75 cm), were considered while the CBs were at their critical positions. The results in Table 4-16 show that the target peak linear power and total assembly power are - kW/m and - kW, respectively, which are less than those of the limiting core cases. The maximum inner and outer plate peaking factors of the diver fuel are 3.043 and 2.576, respectively, for the minimum bumup core. The effect of regulating rod position is relatively small when compared with that of the CB.

Table 4-16. Effect of regulating rod position on target power and peak linear power Regulating Target Peak LP Inner plate Outer plate Core state r~d (cm) power(~W) .(kW/m) . peaking peaking Maximum bumup 38.1 2.637 2.218 with equilibrium 31.75 2.641 2.147 xenon 25.4 2.638 2.046 Average bumup 38.1 2.511 2.155 with equilibrium 31.75 2.542 2.093 xenon 25.4 2.528 1.973 38.1 3.037 2.576 Minimum bumup 31.75 3.043 2.563 without xenon Extreme bumup without xenon 25.4 38.1 31.75 25.4

--- --- 3.031 2.988 2.999 2.976 2.565 2.531 2.507 2.397 4.5

• Uncertainty Analysis The physics analysis has an inherent uncertainty due to the solution method of the MCNP6 code. The standard deviation (1 a) of the pellet power is 1.1 % for almost all the pellet numerical nodes when 100 million particles are used for the calculation. For the total target power, the standard deviation (1a) is as small as 0.26%. Other uncertainties considered are impurities of target uranium, pellet density, fissile content and target rod position due to manufacturing tolerance. Table 4-17 shows impurities of of the target uranium, cladding, cartridge and neutron shield. For the aluminum (cartridge) and stainles steel (neutron shield), the impurity was assumed to be 10 part per million (ppm) natural boron.

62

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 4-17. Material impurities for uncertainty analysis

' Uranium Zircaloy-4 Aluminum Stainless steel (wt%) (wt%) (ppm) (ppm) 232u 2x10-7 234u 0.26 235u 0.46 Boron 3ppm 0.00005 10 10 4.5.1 Sensitivity calculations The uncertainties of the total target power and peak linear power were estimated for for the limiting core configurations of all four core burnup states. In order to estimate the sensitivity of the core and target performance parameters, the uncertainties of design parameters were defined as follows:

• For the statistical uncertainty of the solution method, +/-2a value is used to estimate the eigenvalue, total target power and peak linear power with a 95% confidence level.

• The impurities always reduce the target power and are regarded as a bias. The uncertainty is estimated as 95% of performance parameter change due to the maximum impurity level of the target uranium, assuming a flat distribution of impurity between its minimum and maximum values.

• The manufacturing tolerances of the pellet density and fissile content are

• respectively. The target performance is relatively insensitive to the variation of these parameters. For conservatism, the tolerance value is taken as the uncertainty. In the simulation, only a positive value is used as a bias to estimate the power increases.

• The uncertainty of the target rod position is -

• which is

• of the manufacturing/

installation tolerance -

• i.e., approximately a 2a value when a triangular distribution is assumed between the minimum and maximum values with the mode (average) at 0. The - means that the rod is closer to the core b y -

• Though the target rod could be either closer to or farther from the core, - is used as a bias to estimate the power increase of the target.

The estimated uncertainties (positive components) of the core eigenvalue, target total power and peak linear power due to the statistical and manufacturing uncertainties are summarized in Table 4-18, Table 4-19 and Table 4-20, respectively. The impact of the pellet fabrication density and fissile content uncertainty on the core eigenvalue and target peak linear power is very small, which makes it difficult to obtain consistent results by direct perturbation calculations.

63

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Therefore, these uncertainties were increased by a factor of 5 in the direct perturbation calculation, and the results were linearly interpolated.

The uncertainty (positive variation) of the eigenvalue due to manufacturing tolerance (pellet density and fissile content) is estimated to be less than

• pcm. The impact of rod position uncertainty is the largest among five uncertainties for all core states. For the target total power, the statistical uncertainty is very small, while the target rod position uncertainty dominates the total uncertainty. The estimated uncertainty due to manufacturing tolerance is - and_

for the pellet density and enrichment, respectively, while the uncertainty due to target rod position is - for the maximum bumup core. For the peak linear power, the statistical uncertainty prevails over the uncertainties due to the pellet density and enrichment, but it is comparable to the uncertainty due to target rod position.

Table 4-18. Uncertainty of core eigenvalue due to simulation and manufacturing Statistical Fabrication Rod Core burnup h~1purity Enrichment*

uncertainty density. positio11 ..

state (pcm) (pcm)

(pcm) (pcm) focm)

Maximum with equilibrium xenon

• *

• I Average with equilibrium xenon

• *

• I Minimum without xenon

• I I I Extreme without xenon

• I I I Table 4-19. Uncertainty of target power due to simulation and manufacturing Core burnup state Maximum with equilibrium xenon

. Statistical uncertainty (%)

Impurity (%)

Fabri~tion

• densify.(%)

Enrichment

. (%)

Rod position (%)

Average with equilibrium xenon Minimum without xenon Extreme without xenon 64

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 4-20. Uncertainty of peak linear power due to simulation and manufacturing Fabrication *: **Enrichment Rod Statisticai **

Maximum with eciuilibrium xenon un~ertainfy (%)

Impurity (°lo) dens.ity (%) .

(%) *" .. * , positi6h

. ,*1%1 ..

Average with eciuilibrium xenon Minimum without xenon Extreme without xenon 4.5.2 Total uncertainties of the limiting core performance parameters The total uncertainties of the key performance parameters (eigenvalue, total target power, and peak linear power) were estimated in two ways as summarized in Table 4-21:

• A simple arithmetic summation of individual uncertainties. Note that the uncertainty due to impurity is always negative, bl,Jt is not included in the summation for conservatism.

• A root-mean-square (RMS) of individual uncertainties. The total uncertainty is obtained as a product of statistical uncertainty and RMS of fabrication density, enrichment, and target rod position uncertainties. For example, the upper bound of the power (P) is estimated as P=Po(l +crs) ( 1+ cra+cr~+cr~). where Po is the power without uncertainty, subscripts s, d, e, and prefer to the uncertainties due to statistical, density, enrichment, and position, respectively. The uncertainty due to impurity is neglected for conservatism.

The simple arithmatic summation means that individual constituent uncertainties occur simultaneously, which is a very low probability case and could be too conservative. It was decided to use the RMS model to estimate the total uncertainty. For the limiting core configurations, the total uncertainties were finally applied to the design values of the target performance parameters:

• For the eigenvalue, the estimated uncertainty is -

• For the total target power, the estimated uncertainty is +/-1.74 % for the maximum bumup core. The estimated upper bound of the total target power is then - kW (cf. -

kW without uncertainty).

65

ATTACHMENT .6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B

• For the peak linear power, the estimated uncertainty is - for the extreme burnup core. The estimated upper bound of the peak linear power is then - kW/m (cf. -

kW/m without uncertainty).

Table 4-21. Total uncertainty of the performance parameter Core Eigenvalue (pcm) Target total pow~r (%) Peak linear power (%)

burn up state Arithmetic RMS Arithmetic RMS Arithmetic '

RMS Maximum Average ** **

Minimum Extreme ** ** - - - -

4.6 Target Material Depletion As the fissile material burns in the target assembly, the target assembly power steadily decreases as the target assemblies are being irradiated, under the condition that the incoming neutron flux level is constant. Table 4-22 lists the isotopic mass of two target assemblies irradiated for two weeks for the average bumup core, which is the most probably core state.

The 99Mo production calculations have been conducted under following core conditions:

• Mid-life (- 4-year) Be reflector composition is used.

• The average CB age (4-year) is used for all CB's. There is no CB mechanical tilt.

• The average burnup core fuel composition is used for driver fuel elements. The CB's are at their critical position.

99 The Mo buildup is shown in Figure 4-29 for different loading and operational schemes. For the base operation scheme, the calculated 99 Mo weekly production is - 6-day Curies, When I fresh target assemblies are irradiated for - - hours) and discharged. In order to satisfy the requirement of 3,000 6-day Curies per week, one or both reflector shall be loaded twice per week.

99 For the staggered operation, the weekly Mo production is - 6-day Curies, when 2 target assemblies are irradiated for - in an alternate way. For the partial loading with - per assembly, the calculated weekly production i s

• 6-day Curies. Based on these results, the 99 Mo production per target rod is obtained as 6-day Curies per week for the base, staggered and partial loading pattern, respectively. For the partial loading, the production rate from - r o d per assembly is 6-day Curies, 66

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B respectively, for two-target assembly loading. It should be noted again that these products rates were estimated under the assumption of constant neutron flux in the target which will result in an overestimation of the production rate.

As the fissile uranium depletes, the target assembly power linearly decreases as shown in Figure 4-30. At the end of -

the base loading pattern, which corresponds to a power of . . kW. The average target pellet burnup is -

irradiation, the target assembly power drops to . . kW for reduction when compared to the initial Mega Watt day per ton Heavy Metal (MWd/tHM).

Table 4-22. Target material composition and burn up for the average burnup core Calendar day I

Operation Burn up (MWd/tHM)

I 23su (gram) 23au (gram) 239pu (gram)

I 99 Mo (mg)

I I

Figure 4-29. Estimation of weekly 99Mo production for different loading patterns 67

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00031/B I1 I

I LJ I I

I -

I I

0 2 4 6 8 10 12 14 Operation time (day)

Figure 4-30. Estimation of the target assembly power during - operation 4.7 Structural Component Heating 4.7.1 Steady-state structural component heating The structural component heating was estimated by MCNP6 code using the kinetic energy deposited by the fission fragments, prompt neutrons and the delayed neutrons. In addition to the neutron data used for neutron flux calculation, the heating calculation uses photon-atomic data (mcplib04) [White 2003], photon-nuclear data (endf7u) [White 2000], electron data (el03)

[Briesmeister 2000], and delayed neutron/gamma data (cinder.dat, delay_library.dat, cindergl.dat, delay_library_v2.dat) [Durkee 2012, Pelowitz 2013]. The MCNP6 adopts the CINDER'90 model to calculate delayed particle emissions from all of the radionuclides in the decay chain for a fission product. The photon emission time for delayed-neutron and delayed-photon emission was adjusted to 105 sec (the default value is 1010 sec), under which the statistical uncertainty (1 a) of the calculated heat deposit in the Be reflector is *

• The structural component heating is summarized in Table 4-23 for the maximum burnup core which has the highest radiation heating of the components. The Be reflector heating is higher for the maximum bumup core by ml when compared with the minimum burnup core. For the maximum burnup core, The Be reflector heating is . . kW and - kW with and without two target assemblies, respectively. The heat in the cartridge and neutron shield is - kW and

• kW, respectively, when two target assemblies are loaded.

The spatial distribution of the Be reflector heating is given in Table 4-24 for 6 azimuthal sectors and

• segments of sector I which faces target assemblies. The aegment of sector I 68

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B I I consists of axial nodes and radial nodes. Though the total Be reflector heating is the highest for the maximum burnup core, the local heating is the highest for the minimum burnup core, when the CB is at its estimated lowest position, due to skewed axial power shape as summarized in Table 4-25.

Table 4-23. Component radiation heating of the maximum burnup core Target Neutron Photon Electron Component Total (kW) loading heating heating heating No target Be reflector Be reflector Two target AI cartridge assemblies Steel shield Table 4-24. Spatial distribution of Be reflector radiation heating for the maximum burnup core Target Heating Axial layer Radial segmentation of Sector 1 Sector loading (kW) of sector 1 Inner Middle Outer I I I I No target I

I I -..- I I

I ** ** *-

I I

I -..- I I -*- -*- -*-

Two target assemblies I

I I

I

-..- I I

I -- -- --

69

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 4-25. Spatial distribution of Be reflector radiation heating for the minimum burnup core Target Heating Axial layer Radial segmentation of Sector 1 Sector loading (kW) of sector 1 Inner Middle Outer I I I I No target I

I I --- I I

I -- -- --

I I

I --- I I --- --- ---

Two target assemblies I

I I

I

--- I I

I -- -- --

4. 7.2 Post-shutdown structural component heating The post-shutdown structural component heating was estimated by a combination of ORIGEN2 and MCNP6 calculations. This is necessary due to photo-neutron effects after shutdown, where photons from MURR driver fuel and SGE target rods react with the Be reflector, releasing neutrons which then cause fission in the SGE target rods, leading to heating in rods and assembly components.

ORIGEN2 cases modeled irradiation of the MURR driver fuel elements to their respective burnup values, followed by several decay steps. Results show that the initial

• minutes into the decay has the largest reduction in photon source (gammas released per second). Details of the photon sources (18-group mean energy and probability distributions) were input to MCNP6 via the SDEF (source definition) card. Photonuclear physics was turned on by the MODE card. The same nuclear, photon-atomic, photon-nuclear, and electron data was utilized as the steady-state structural component heating calculations.

Tallied results indicate that the total heating is very small. One second into the decay, total heating is - W for the beryllium reflector, - W for the SGE target rods and . . W for the SGE target assemblies. After *minutes into the decay, total heating reduces to - W for the beryllium reflector, - W for the SGE target rods and - W for the SGE target assemblies.

70

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B These results are considered credible upon closer evaluation. The total photon (p) flux in the beryllium reflector is - p/cm2-sec. Noting that the threshold for photonuclear reactions are for energies greater than 1.6 MeV, the photon flux in the beryllium reflector within the threshold is reduced to - p/cm2-sec. The total photonuclear cross section (from the '?Ou' library) is very small, averaging 4.3x104 barns. Using these numbers, the photonuclear reaction rate in the Be reflector is - reactions/cm 3-sec, essentially the initial neutron density in the Be reflector. This low neutron density made it difficult to achieve great statistics, where the tallied target rod neutron flux is - n/cm2-sec with an error of 15%. These results confirm that post-shutdown photo-neutron effects have a negligible impact on heating in the Be reflector, SGE target rods and SGE assembly components.

5 VALIDATION OF METHODS USED IN PHYSICS DESIGN It is essential that the computational tools and methods should predict the target assembly as well as the reactor core accurately so that the target assembly operates safely and economically as designed. The confidence in the nuclear design and analysis results can only be obtained by comparing calculated results with measurement data. The MCNP code has been benchmarked against various experimental results [Whalen 1992, Mosteller 2002, Mosteller 201 O] and recognized as the most robust tool in criticality and radiation-shielding calculations. However, the code validation is an integral assessment of the solution method, geometry modeling, and nuclear data and, therefore, it is problem-specific. The best benchmark model of MCNP for the application to the target assembly design and analysis would be validating MURR core itself or MURR-like core by the MCNP code. Two benchmark models of MCNP and one verification test of the 99 Mo production model have been selected as follow:

• Section 5.1 presents MURR technical report on the validation of MCNP model for predicting critical control blade position. The model includes detailed fuel and core geometry as well as depletion of fuel and structural components such as control blades and beryllium reflector [Peters 2013].

• Section 5.2 presents the validation results of MURR control blade depletion model. The model includes detailed segmentation of the control blade based on their in-core service period and comparison to measured criticality [Peters 2012a].

• Section 5.3 presents a criticality benchmarking of ATR [Kim 2005], of which the fuel specifications are similar to those of MURR. The core is also cooled by light water and reflected by beryllium.

• Section 5.4 presents a verification test of the 99Mo prediction model against the ORIGEN2.2 code for 99Mo content in the target assembly and collection system

[Choi 2015b].

71

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 5.1 Validation for Computational Methods and Calculations at MURR MURR Core Description and MCNP Model The MURR has a cylindrical core consisting of eight highly-enriched uranium (HEU) fuel elements surrounding a central flux trap. A primary beryllium reflector ring is located inside a secondary graphite reflector ring, which surrounds the core. The core is cooled and moderated by light water. Several irradiation locations are present in the flux trap region as well as the graphite reflector region of the core. Its excess reactivity is controlled by the movement of four axially-banked Boral shim control rods (blades). The blades travel within an annular water gap between the outer surface of the outer pressure vessel and the inner surface of the beryllium reflector. Figure 5-1 shows the MCNP model of the MURR core configuration .

Figure 5-1. A detailed MCNP model of the MURR core To properly model a MURR's "mixed-core" fuel configuration in MCNP, a separate fuel depletion study was done previously using MONTEBURNS-2.0 [Trellue 1999]. This study generated the fuel material definitions needed in MCNP, i.e., atom densities of various isotopes in the fuel matrix, for a range of xenon-free fuel elements from fresh (0 MW-days or no fuel depletion) to almost spent status (approximately 150 MW-days). For the ECP prediction calculation, fuel 72

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B element definitions can be individually selected from this fuel burn up database to make up the starting xenon-free core comprising of eight fuel elements.

In addition to having a mixed-core fuel configuration, MURR also employs a mixture of four independently depleted Boral shim control blades (rods). Here, each blade has a different core runtime, and hence, different axial and radial boron depletion profile. In order to accurately predict the ECP, the depletion state of each in-core shim control blades have to be accurately defined in the MCNP core model. Similar to the fuel depletion simulations, a separate control blade depletion study [Peters 2012a] was also undertaken to generate control blade definitions for various stages of control blade burn up.

Other major factors that influence the ECP prediction , which were included in the MCNP model are the runtime depletion of the beryllium reflector and the status of the central flux trap region.

Depletion studies of the beryllium reflector using MONTEBURNS were done to understand the changes in the core's reactivity as a function of runtime. This study produced a library of beryllium material definitions describing its burnup status at various runtimes from a fresh state up to 8 years; MURR usually replaces the beryllium reflector after every 8 years of runtime. As with the fuel and control rods databases, a particular beryllium material definition can be selected from the library to improve the accuracy of the full-core MCNP model. MURR utilizes the central flux trap, the peak thermal flux region of the core, extensively for irradiating various positive and negative reactivity-worth samples each week for research as well as commercial applications. The mixtures of samples irradiated in this region vary slightly from week to week.

Reactor Safety Limits also limits the total reactivity worth of the flux trap. Therefore, in the MCNP model, various targets that go into the flux trap region as well as the graphite reflector of the core were modeled accurately in order to reduce the error in the ECP prediction.

Using the information previously discussed the final MCNP MURR model was created to accurately reflect the status of the fuel , control rod beryllium reflector and flux trap or a given core configuration. The MCNPS KCODE option was used for all calculations. The parameters were set to run 20 million particles, (i.e., 20000 particles per 1000 active cycles). The (n,p)

MODE of calculations was selected and the ENDF-BVll.O neutron cross-section data set was used for both fast and (s,a) thermal scattering neutron interactions. Specifically, the (s,a) thermal scattering data for all moderator material was initially set at 300K.

Automated Critical Rod Height Search Routine A critical rod height search routine was developed to predict the ECP for a given weekly xenon-free MURR core configuration. Given an initial guess of the critical rod height, the search routine utilizes a series of MCNPS criticality (KCODE) calculations in order to calculate the critical control rod height. The routine first uses the initial critical rod height guess in a KCODE calculation to estimate the ke!f for the selected MURR core configuration. If the calculated 73

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B keff ¢1 .0000 (within a specified error range), the routine will automatically adjust the control rod height in a subsequent KCODE calculation . The process is repeated until the predicted keff is essentially equal to 1.0000. For this work, the accepted tolerance for the reactor considered being critical was set at 1.0000+/-0.03%.

Modified MONTEBURNS To accurately predict the hot-startup ECPs, it is necessary to track the buildup of xenon in the core during start up and subsequent steady state operation as well as the buildup and decay of the poisons during shutdown and restart. MONTEBURNS 2.0 was selected to simulate the fission products poison transients (particularly Xe-135) and the ECP prediction of the MURR core.

MONTEBURNS utilizes the capabilities of ORIGEN 2.2 for isotope generation and depletion calculations and MCNP5 for continuous energy, flux and reaction rate as well as criticality calculations. However, the program is not designed as such to handle the transient period from reactor startup through critical, and into steady state operation without some modifications since the process involves the control rod motion as well. Although there are computer codes available that can track control rod movements, it was decided to modify and utilize the system of codes that are already adapted to MURR core modeling and predictions. Therefore, MURR Hot Startup ECP predictions were performed using a modified MONTEBURNS by incorporating the in-house developed critical rod search routine. The data-flow diagram for the modified version of MONTEBURNS used here is shown in Figure 5-2.

74

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B MONIEBURNS 2.0 ORIGEN 2.2 MCNP5 Mass Detailed Fluxes & Activity Reaction Rates { Heatload Burn up Critical Rod Search Critical Rod Height Figure 5-2. A depiction of the generalized information flow in the modified MONTEBURNS simulation for predicting hot startup ECPs The normal flow of data in MONTEBURNS is a cyclic, stepwise process where every depletion or burn step involves both MCNP and ORIGEN calculations, and every decay step employs ORIGEN calculations only. In Figure 5-2, this is represented as the flow between the MCNP5 and ORIGEN2.2 within the MONTEBURNS block. In the modified MONTEBURNS simulations, the Critical Rod Search routine is placed external to this scheme and linked as shown in Figure 5-2. The Critical Rod Search routine ensures that the MCNP5 model that is used in the depletion calculation for a particular time-step in MONTEBURNS is that which represents the critical core state. This modified MONTEBURNS simulation methodology allows for the study of the MURR-core at any state-point; that is, whether at xenon-free initial critical state, or at the end-of-cycle equilibrium-xenon critical state or at other specified point in between. Moreover, when the fuel 'feed' option is employed in MONTEBURNS, various startup and shutdown scenarios of the MURR core can be simulated with high accuracy.

Benchmarking ECP Predictions from the Computational Methodology In order for the shutdown scenarios to be accurately simulated , the results from the critical rod height search routine were first benchmarked separately. Here, initial ECP predictions using the critical rod search routine for various weekly startup cores were compared with the actual initial (start-up) critical data. Eight weeks of benchmarking were undertaken to ensure consistency in the routine's ability to predict the initial ECP accurately. Table 5-1 shows a span of data 75

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B reported over eight months. Startups at the MURR require an occasional "strainer" start up where initial critical data is taken without any flux-trap samples or irradiation tubes. Two strainer startups were reported in Table 5-1 .

Table 5-1. A comparison of predicted initial critical rod height vs. actual at the MURR Core Configuration Actual Initial Predicted Initial Predicted Keffective Flux Trap Date Critical Height (inches) Critical Height (inches) Configuration Week 1/28/2013 16.79 16.67 0.99993 Strainer Week 2/04/2013 16.52 16.27 0.99975 FT Samples Week 4/29/2013 15.98 15.78 1.00017 FT Samples Week 6/10/2013 15.44 15.42 0.99995 FT Samples Week 8/05/2013 16.74 16.74 0.99985 Strainer Week 8/12/2013 15.71 15.61 0.99985 FT samples Week 8/19/2013 15.84 15.84 1.00016 FT samples Week 8/26/2013 15.64 15.69 1.00029 FT samples A negative bias of -1 .5% is seen in the predictions for the early benchmarks for the various core configurations. After some additional refinement of the MURR core MCNP model, which included the additional effects of MURR's aged beryllium reflector ring, and more appropriate (s,a) thermal scattering data for neutron in the primary coolant (i.e., data at 350K vs . 300K), the variations in the predictions were within +/-0.8% of the actual critical rod heights. Figure 5-3 shows a plot of the predicted ECP percent deviation from actual critical. The improvements based on adding the beryllium age-effect and the higher-temperature (s,a) tables in the primary coolant/moderator to the MCNP models are both indicative of adding negative reactivity to the core, hence causing an increase in the control rod heights to values closer to the actual values.

76

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B ECP Percent Deviation from Actual Critical 0.50%

0.00%

I 1 2 3 ~

• 5 6

• 7 8 Week

• ECP Percent Deviation from Actuol Critical

-1.50%

-2.00%

Figure 5-3. A plot of the predicted ECP vs. actual initial critical height The first hot-startup benchmark was based on an unscheduled shutdown event occurred during normal operation initially starting with the week-4 cold clean core configuration having total megawatt-days of 600.9 and an average control rod runtime of 4 years. Here, the unscheduled shutdown occurred 11.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after normal operation at 1O MW following the initial startup. The total down period of the reactor, which includes to where the reactor is critical at 50 kW, lasted for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Here, the low-powered Hot-start criticality (i.e., reactor power less than 50 kW) was achieved in 19 minutes and to a power of 5 MW at 24 minutes after startup. Based on this information, the event was simulated in this work and the results are presented in Table 5-2.

Another occurrence of an actual hot-startup attempt was simulated , where an unscheduled shutdown occurred 26.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after operating at full power following the initial startup, using a core with total megawatt-days of 613.7 and an average control rod runtime of 3.625 years. The hot-startup was attempted after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of downtime. Failure for the reactor to restart indicated that the xenon-135 concentration could not be override by the withdrawal of the control rods even at maximum travel. The benchmark of this simulation is presented in Table 5-3.

77

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 5-2. A Comparison of the predicted ECP vs recorded rod height data for an actual hot startup from an unscheduled-shutdown event Time into normal operation Actual Control Rod height Predicted Control Rod Height  % Deviation from critical (Hours) inches inches 0 15.44 15.41 0.194%

5.75 17.16 17.19 -0.175%

11.5 19.25 19.15 0.519%

Time into Hot Startup operation from critical (Hours) 0 22.95 22.73 0.959%

1.00 21.04 21.00 0.190%

The data presented in Table 5-2 shows very good agreement between the predicted control-rod heights and the actual recorded heights at startup, during normal operations and at shutdown, after 11.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The deviations between the predictions and actual control-rod heights are less than +/-1.0%. At MURR, almost all of the excess-reactivity loss during normal operation is due to xenon poisoning, therefore it is seen here that the transient xenon-135 quantity is predicted accurately at any given state point. Since Hot-startups rely on whether there is sufficient core excess-reactivity established to override the buildup of xenon-135 by continuous withdrawal of the rods, taking startup critical rod-height data during hot-startups is not as precise as a normal startup and is often not recorded. For this simulated unscheduled scenario, the predicted hot-startup ECP is quite close to the observed startup rod height of 22.95 inches. At one hour of operation after the hot-startup, the simulation predicts a lowering in the control rods. This response is expected, and is seen in the actual data since the xenon-135 burnout rate at this point is larger than its buildup rate from the decay of iodine-135, causing a decrease in the total xenon-135 concentration, therefore, adding positive reactivity to the core .

Table 5-3. A Comparison of the predicted ECP vs recorded rod height data for an actual failed hot startup from an unscheduled-shutdown event Time into normal operation Actual Control Rod height Predicted Control Rod Height  % Deviation from critical (Hours) inches inches First startup with loaded FT 0 16.25 16.26 -0.062%

26.00 22.8 22.63 0.746%

Time into Hot Startup operation from critical (Hours) 0.0833 no startup 26.00 Maximum CR travel Maximum CR travel 78

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B For the next unplanned shutdown simulation , Table 5-3 shows the control rod height comparison between the simulation and the actual measurements for this hot-startup attempt.

The deviations between the predicted critical rod heights and the actual recorded rod heights show good agreement i.e., within 1.0%. Like the actual hot-startup attempt, the simulation predicts no possibility for a hot-startup; here, the control rods are fully withdrawn (at 26.0 inches) with a predicted subcritical ke11 of 0.99568 .

5.2 Validation of MURR Control Blade Depletion Model The University of Missouri Research Reactor (MURR) is a 10 MW research and training reactor that is owned and operated by the University of Missouri-Columbia . The reactor is designed as a compact-type , pressurized , cylindrical core consisting of eight fuel elements surrounding a central flux trap. The core is primarily moderated by light water and has a primary beryllium reflector inside a secondary graphite reflector ring. Controlling excess reactivity for reactor operation comes from the movement of four axially-banked BORAL shim control blades. The blades travel within an annular water channel between the outer surface of the outer reactor pressure vessel and the inner wall of the beryllium reflector. Figure 5-4 below shows a horizontal cross-sectional view of the MURR core. The poison zone of each blade spans an arc of 72 degrees, has a length of 34 inches and a thickness of 0.1 inch . The travel distance for the active region of each blade is constrained to the length of the fuel meat at 26 inches. A fifth control blade, constructed of stainless steel and located in the same water channel , is used for only very minor adjustments to excess reactivity during steady-state full power operation .

Figure 5-4. Cross sectional view of the MURR core 79

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B In collaboration with the Global Threat Reduction Initiative (GTRI) program at Argonne National Laboratory (ANL), MURR took part in a feasibility study [McKibben 2009] for the use of a low-enriched uranium (LEU) molybdenum alloy (U-10Mo) as a new monolithic fuel in the reactor core. The neutronic analyses performed for the feasibility study included a criticality study using LEU fuel, a fuel depletion analysis, as well as a powe~ peaking analysis for an all-fresh core and for mixed cores consisting of fuel elements at various stages of burnup. In order to benchmark the feasibility neutronic analyses, which were performed using the 3-dimensional radiation transport code MCNP [LANL 2003] and ANL's transport and depletion modules DIF3D

[Lawrence 1983] and REBUS [Olson 2001 ], several criticality analyses were performed for the current MURR highly-enriched uranium (HEU) core. Although the criticality benchmarks predicted kett values relatively close to the measured values, a deficiency was noted where all of the studies employed fresh (non-depleted) control blades. This tended to bias the predicted keff values.

The use of control blades in the MURR core is very complex since several "active" shim control blades are cycled in and out of the core multiple times during their useable lifetime and, at any given time, blades in each of the four possible control blade locations are at different stages of burnup. Currently, there are 19 active control blades in use at the MURR. These active control blades have an independent operational cycling scheme where each blade has about a two-year irradiation (or core residence) time in one of four positions - "A", "B", "C" or "D" - of the MURR core (see Figure 5-4) and a potential maximum lifespan of about 10 years - providing the blade is usable after each two-year irradiation cycle. Since a given replacement blade is selected with minimal consideration given to its burnup status, for a set of in-cycle blades, the "effective" lengths and atom density of boron-10 (108) in the poison zone for each blade may vary significantly. This disproportion in the effective lengths of the poison regions between the blades is expected to affect the power peaking factors on each fuel element differently. Although the MURR HEU fuel cycle and predictions of the LEU fuel cycle were studied in detail using the ANL-MURR MCNP models in the LEU feasibility study [McKibben 2009], until now no similar study for the control blades was available.

The importance of using a full depletion model of the control blades for the reactor safety analysis is clear. In this work, a methodology was developed to individually model the 108 depletion history for each active control blade that went through the MURR control blade operational cycle. Verification of this methodology was done by benchmarking the calculated keff. using control blade depletion models against actual critical rod height measurements.

80

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Computational Methodology MCNP Model Highly discretized control blade models representing each of four possible positions "A", "B", "C" and "D" of the MURR core configuration were incorporated into the ANL-MURR MCNP model.

Each control blade model was divided into 292 independent zones (cells), which include:

• 4 azimuthal sections (over the entire length)

• 9 radial sections over each 1/4 inch sections of the first axial (bottom) inch for each azimuthal section

• 9 radial sections over each of the next three incremental one-inch sections for each azimuthal section

• 2 radial sections over each of the next five incremental six-inch sections for each azimuthal section This discretization scheme was chosen to properly account for the 108 atom density burnout profile in the most active regions of each control blade. Figure 5-5 shows the increased discretization towards the bottom tip of the control blade model. Because the bottom portion of any control blade is constrained to be within (or near) the length of the fuel meat zone whether at startup, at equilibrium xenon build-up, or at the almost fully withdrawn state, it is expected to have the fastest rate of 108 depletion.

81

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Figure 5-5. Axial and radial discretization in the bottom 4nof the MURR MCNP control blade model Depletion Methodology 10 The 8 depletion calculations were done using the actual run-time histories for each of the 19 active control blades, shuffling them through their life-cycle positions "A", "B", "C" or "D" as appropriate. Each control blade initially begins its potential ten-year cycle with the atom densities for fresh BORAL in each depletion zone. To include control blade travel during the 150-hour weekly operating cycle , the fuel definitions for two state points of the typical HEU mixed core cycle were used for improved accuracy; i.e., fuel definitions for the control blade (CB) at 17 inches and CB at 23 inches. Here it was approximated that the reactor configuration at the first state point (i.e ., CB height at 17 inches) accounts for the first 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of the typical MURR one week-long operating cycle (where the xenon is building up in the core and the blades are moving out) . The second state point (i.e., CB height at 23 inches) accounts for the remaining 114 hours0.00132 days <br />0.0317 hours <br />1.884921e-4 weeks <br />4.3377e-5 months <br /> of the weekly operating cycle (where equilibrium xenon levels are established and the shim control blades are mostly out of the peak flux reg ion of the core) .

10 The B (n ,a) reaction rates in each zone and for each state point were tallied in MCNP. Next, 10 the reaction rates were used in a macro-modified EXCELc worksheet where the actual B atom density depletion in each zone was calculated . For the control blade burnup simulation , a time 82

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B step of six months was used and each control blade was irradiated for a minimum of two years before being set aside for decay and potential reuse.

Results Effects of Core Configuration on Control Blades Depletion 10 To study the effects o.f different reactor configurations on the B (n ,a) reaction rate profiles in 10 each control blade position , the B (n , a) reaction rate in each zone of each blade position was tallied and compared for fresh blades in eight different reactor configurations. Various reactor configurations were simulated using the ANL-MURR MCNP models to reflect the major changes in the graphite reflector region over a 30-year period . The 1971 reflector was nearly a single sleeve of graphite surrounding the core while the 2008 graphite region possesses much less graphite due to the addition of numerous experimental irradiation channels. The material definitions for the fuel assemblies in the weekly start-up core were chosen from two bumup states; either all-fresh fuel or a typical cycle consisting of both fresh and partially depleted fuel assemblies derived from an ANL REBUS-DIF3D simulation of the MURR HEU fuel cycle. The control blade positions for each weekly operational cycle were either approximate critical blade height at startup or approximate height at equilibrium xenon levels. The different reactor configurations used are as follows:

(a) all fresh fuel , 1971 graphite reflector; critical (CB at 17 inches)

(b) all fresh fuel , 1971 graphite reflector; equilibrium Xe (CB at 23 inches)

(c) all fresh fuel , 2008 graphite reflector; critical (CB at 17 inches)

(d) all fresh fuel , 2008 graphite reflector; equilibrium Xe (CB at 23 inches)

(e) mixed fuel , 1971 graphite reflector; critical (CB at 17 inches)

(f) mixed fuel , 1971 graphite reflector; equilibrium Xe (CB at 23 inches)

(g) mixed fuel , 2008 graphite reflector; critical (CB at 17 inches)

(h) mixed fuel, 2008 graphite reflector; equilibrium Xe (CB at 23 inches) 10 Plots of the B (n,a) reaction rates vs. the axial and radial dimensions for each control blade depletion zone were examined and showed that each reactor configuration has a unique impact 10 on the B (n,a) reaction rate profiles in each blade position. However, the effects were observed to be small and are therefore not very significant. The results also showed that each blade position has a unique profile and that the most active regions of the blade are within the bottom four inches of the blade. Figures 5-6 and 5-7 show examples of the combined axial and 10 radial B (n,a) reaction rate profiles over the blade's bottom four inches, for reactor 83

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B configurations (e) and (g) - MURR's current typical configuration. It also shows that the highest reaction rates occur at the surfaces of the control blade.

9.00E.otl ~. 1/4 ftd\ of C1I Olefl ldp 8.00E.otl

~

"ii 7.00[.()8 6.00E.otl -

Secood 1/4 iftdl of ct deOl.eft Th d 1/4Mdlof ctl . . . Oleoll

!Ill S.OOE-08 Edee v - fourth l/4 lndl ol Ji 4 .OOE-08 Cl de 0 l.eft II [

~ 3.00E-08 ~ llndlofct c 111.ade o ult £

-11 2.00[.()8

.... 1 lndl of C1I

~ 1.00E.()8 111.ade ldt (

O.OOE*OO J.-u.11ncllolc.1

.0.060 .0.040 .().020 o,ooo 0.020 0.040 0.060 l [dee Rad~I thickness of 81~ (Inches) 10 Figure 5-6. Control blade B reaction rates for; mixed fuel core; 1971 graphite reflector; critical (CB at 17 inches) 9.00E.08 -.....-- - - - - - - - - - - - - - - - - - .

- First 1/4 Inch of CB 8.00E-08

• 1--- - - - - - - - - - - - - ---- --1 Blade 0 left Ed1e

=

1,,

.. 7.00E*08 -- - Second 1/4 Inch of CB Blade D left edae

~ 6.00E-OS Ill - Third 1/4 Inch of CB v

.!: S.OOE-OS Blade D left Edae

~

CIC 4 .00E*08 +-- -,a*a -- - - - - - - - - --il - Fourth 1/4 Inch of CB c Blade D left Edae

~ 3.00E*OS

.....£ 2.00E*08 - Second 1 Inch of CB Blade D Left Edae 1.00E*08 Third 1 Inch of CB Blade D Left Edae O.OOE+OO .i---------------~--~

• 0.060 .().040 -0.020 0.000 0.020 0.040 o. fourth 1 Inch of CB Radial thickness of Blade (inches) Blade O left Ed1e 10 Figure 5-7. Control blade B reaction rates for; mixed fuel core; 2008 graphite reflector; critical (CB at 17 inches) 84

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Benchmarking: Comparison of Calculated ~ Values The data presented in Table 3.8 of Ref. [McKibben 2009] (LEU feasibility study) showed the deviation of calculated core kerJ values for the measured critical control blade positions for various blade histories. For that study, the partially burned fuel assemblies for each criticality measurement was modeled using fuel compositions generated by a REBUS-DIF3D simulation of a two-year HEU fuel cycle shuffling scheme done for MURR in previous work [McKibben 2009; Deen 2003]. The average burnup of the cores was 620 Megawatt-days and all of the control blades were modeled as fresh (non-depleted). The flux trap content was modeled as the typically loaded configuration and the graphite reflector was modeled as the "2008" configuration. Several cores from that previous study were selected and the core keff values were recalculated with depleted control blade models during the present study.

Table 5-4 shows the deviation in %flk/k from critical state when there is no control blade burnup effect included (as in Ref. [McKibben 2009]) and when there is blade burnup effect included for each in-use control blade (from this work) . When all-fresh control blades were used in the MCNP simulation (the column labeled "No CB Burnup" in Table 5-4), cores with the combination of blades that have the greatest total burnup, and correspondingly the lowest measured critical blade heights, had the greatest deviation from critical at measured critical rod heights. However, when the blades are modeled with the depleted blade compositions (the column labeled "With CB Bumup" in Table 5-4) , the deviation from critical is always smaller and for the case with the maximum control blade history the deviation is reduced from 1.735% to 0.422%.

Table 5-4. Deviations for the estimated kett values from critical for the case with fresh control blades (from feasibility study) and with depleted blades CB History CB Bank No CB Burnup With CB Burnup (In-cycle Days) Height (inches) (% ~k/k) (% ~k/k) 287 17.63 -0.260 -0.232%

308 18.06 -0 .144 -0.139%

1040 17.22 -1.301% -0.730%

1192 16.72 -1 .307% -0.532%

1709 16.64 -1 .743% -0.390%

1835 16.00 -1 .735% -0.422%

Figure 5-8 shows plots of the deviation of the calculated keff values from critical at the measured critical control blade height when the blades are modeled as fresh as well as with their burnup history. As seen from the plot, the result of adding the depleted control blades to the existing HEU MURR core model shows a significant improvement; the average deviation from critical is -0.41 % flk/k when the blade de ared with -1 .1% flk/k if the blades are 85

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B modeled as fresh. It can also be seen from the plot that for those cases with small total control blade burnup history, the two deviations are practically the same - indicating the effect is more pronounced for the cases involving control blades with large total burn up.

However, there is still a small negative bias in the predicted kett values even when the control blade depletion effect is incorporated in the present HEU MURR model. The bias is seen as the slight negative slope in the trend line fitting the solid red squares in Figure 5-8 and could be due 10 to under-depletion of the 8 in some of the larger axial zones above the blade tip where smaller divisions could have been more appropriate. Another contributor to this negative bias could be the MCNP MURR model itself or the bias in the partially burned fuel assemblies incorporated in the weekly start-up cores.

0.500%

0.000%

~

~

-0.500%

.¥

<l

~-1.000%

0 0

-1.500% .

• HEU Cores CB Burn-up (%41</k) 0 HEU Cores no CB Burn (%Ak/k) 0 0

-2.000%

0

- Linear (HEU Cores CB Burn-up (%Ak/k))

200 400 600 800 1000 1200 1400 1600 1800 2000 In-Cycle Days Figure 5-8. kerr deviations from critical vs. control blade burnup history for fresh control blades and those blades with depletion history modeled Blade Byrnoyt Profiles 10 The 8 depletion results for control blade number 5-08, which began operation in position "C" for its first cycle and remained in the same position for two more consecutive cycles, are shown 10 in Figures 5-9, 5-10 and 5-11 . In Figures 5-9 and 5-10, the azimuthally-averaged radial 8 atom densities are shown for the bottom %-inch of the control blade at each time step. This shows 10 that the overall control blade 8 depletions are predominantly a surface driven effect with an asymmetry in the depletion rate on the surface towards the beryllium reflector. This is due to an enhancement in the thermal neutron density on the beryllium side of the blade.

86

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Profile for 1" 1/4 inch -+-omonths E 0.0120 -. . - - - - - - - - - - - - - - - - - - - - -.....

~

.~ 0.0100

• * * * * * * - Gmonths

.......12months

~ -*-18 months

...~ 0.0080 .....,_24 months

..,._30months

?;' 0.0060 ...,_36rronths

• u; c - 42 months

~ 0.0040 - 4Srronths E -+-s4rronths 0

70 0.0020 - GO months

~ 0.0000 L~~~~~~~~~s~~~~~~~;~J 66 months 72 months 5 fl'lils tiulJYls. lo 1111/s 20 1111/s 20 triils 20 trills lO ltlils 5 ltlils 5 fl'lils (8 e Side} e renecto

' side)

Figure 5-9. Bottom half-inch radial depletion profiles for control blade number 5-08 Profile for 2"" 1/4 inch up 0 .0120 - - - - - - - - - - - - - - - - - - - - - - - -..... -+-om0nths E * *

~ 0.0100 ~------=-...1!!!!!!!~~-. .;::--- - - - - -_j ..,._ 12 months

• * - 6months I

e

~0.0080 ~--~:...._--,.~-~;:_ _ __::""""~-31,,~-....;::oi-----I

-+- 18 months

~ 24 months

~ ..... 30month 0

~0 .0060 +----:.,_-~,_,_~'---~t-~.--'lr.:-~r---"~ 36months Z!' - -42 month' v;

~ 0.0040 - 48 months "O -+- 54 months

...~ 0 .0020 CV

- 60months 66 month 0

til 0.0000 72 months s ftli/3 , .. _

ucr11n Cctor Side)

Figure 5-10. Bottom half-inch radial depletion profiles for control blade number 5-08 Figure 5-11 shows the axial profile of the 1OB atom density averaged over radial and azimuthal zones for control blade number 5-08 at residence times of 0.5, 4 .0 and 8.0 years. At 8 years, the 10 8 atom density is essentially zero in the bottom four inches of the blade.

87

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 0.012 - - - - - - - - - - - - - - - - - - - - - - - -

-+-o.s years of residence

- 4.0 years of residence

...... 8.0 years of residence 0

0.00 5.00 10.00 15.00 20.00 25.00 30.00 35.00 Average Distances of Burn Zones (Inches)

Figure 5-11 . Axial 108 atom density profiles for different residence times for control blade number5-08 Conclusion A comprehensive study was performed first to estimate the MURR control blade depletion profile and then to incorporate the depleted control blade models into the existing HEU MURR MCNP core model. Using a typical MURR core configuration in two fully modeled state-points, the full control blade depletion cycle was simulated , which yielded the full burnup history for 19 individual control blades. The initial part of the study focused on understanding the impact of 10 different core configurations on the 8 reaction rate profiles for each blade position. The effect 10 of core configuration was found to have little impact on the 810 depletion. The overall 8 depletion is predominantly a surface driven effect in the bottom four inches of the control blade with an asymmetry in the depletion rate on the surface towards the beryllium reflector.

Combining the depleted blade models with that of partially burned fuel elements for weekly start-up cores significantly improved agreement of the calculated core kett compared to critical measurements. The deviation (%ak/k) of the calculated core ke11 values from critical was reduced from an average of -1 .1%, when the blades are modeled as fresh , to -0.41 % when the depleted blade model is included. There remains a small negative bias in the results, which can be attributed to either the need for finer discretization of the blade depletion model or a bias in the MURR MCNP model developed for the feasibility study.

88

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B 5.3 Advanced Test Reactor Criticality Benchmark Problem The ATR benchmark model is given in the International Handbook of Evaluated Reactor Physics Benchmark Experiment under the category of HEU fuel , metal fuel, and thermal neutron with an identification number HEU-MET-THERM-022 (or ATR-FUND-RESR-001). The ATR-FUND-RESR-001 was initially reviewed and approved for publication by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and was approved by the International Reactor Physics Experiment Evaluation Project (IRPhEP) in 2008 [NEA 2010].

Overview of Experiment The ATR located at the Idaho National Laboratory (INL) is a 250-MW (thermal) high flux test reactor. The ATR critical experiment reported in this evaluation was performed in 1994 (designated as Cycle 103A-2) as part of nuclear requalification testing following the core internals change-out outage. During this outage, core internal components, including the beryllium reflector, were replaced. The ATR core contains 40 fuel elements arranged in a serpentine annulus between and around nine flux traps. The fuel element consists of 19 curved plates of different width, attached to side plates, forming a 45-degree sector of a circular annulus in cross section. The fuel meat consists of highly enriched (93 wt.%) uranium aluminide fuel powder dispersed in aluminum. The fuel plates are moderated by light water, and reflected by beryllium blocks. Table 5-5 compares key core characteristics of MURR [Foyto 2014) and ATR [Marshall 2013]: the fuel type, enrichment, coolant and reflector. The fuel element configuration of MURR and ATR are compared in Figure 5-12. The ATR core model is shown in Figure 5-13.

Table 5-5. Comparison of MURR and ATR core characteristics MURR ATR Thermal power (MW) 10 250 Fuel assemblies in core 8 40 Fuel type U-Alx plate, 24/assembly U-AI plate, 19/assembly 235 235 Fuel material HEU (93% U) HEU (93% U)

Coolant Light water Light water Pressure 68.4 psia 355 psig Reflector Beryllium, Graphite Beryllium 89

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R0003 1/B (a) MURR fuel element 0.073 0.078 w Plate 19

.cs* Nominal Actl Fuel Length 49S Fu Plat Length (b) A TR fuel element Figure 5-12. Comparison of MURR and A TR fuel elements 90

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Figure 5-13. MCNP model of A TR core Evaluation of Experimental Data All the MCNP calculations used continuous-energy cross sections, employing 1250 generations of neutrons with 5,000 histories per generation. The first 50 generations were excluded from the statistics for each case, producing 6,000,000 active histories in each calculation . The statistical uncertainties associated with those Monte Carlo calculations are about +/-0.00035 for the perturbed and the reference cases. The study judged that the sensitivity effect is insignificant when the calculated value is within the Monte Carlo uncertainty (+/-0.00035).

91

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Results of Sample Calculations The results in the International Handbook of Evaluated Reactor Physics Benchmark Experiment are presented in Table 5-6. The evaluated continuous energy ENDF/B-V cross-section data used in MCNP is for 27°C. Additional cross-section data sets have been implemented to provide a comparison. Cross section data for thermal scattering treatment in the JENDL-3.3 library were unavailable so thermal scattering cross sections from the ENDF/B-Vll.1 library were used.

GA has conducted the criticality benchmark calculation of ATR using MCNP6 with ENDF/B-Vll.1 cross section data. A total of 1,000,000 source particles were used: 100,000 particles per cycle and 1000 cycles. The calculation was performed on the Blue High Performance Computing (HPC) Cluster (named "Northshore") at GA. The results are included in Table 5-6 and compared to other.published numbers. The calculated off-criticality is 0.181%Lik which is less than the averaged off-criticality of other five cases (i.e., 0.246%ak).

Table 5-6. ATR criticality benchmark calculation International Handbook of Evaluated Criticality Safety Benchmark GA Exper,i~ents MCNP5 "MCNP5 MCNP5 MCNP5 MCNP5 .,MCNP6 '

(ENDFIB- (END FIB-(ENF/B-Vll.O) (JEFF-3.1) (JENDL-3.3) (ENDF/B-Vll.1)

V.0). Vl.8) 0.99819 +/-

0.9983 0.9937 1.0008 0.9983 1.0018 0.00026 5.4 Verification of Physics Method for Estimating 99Mo Production An analytic physics model was developed to estimate the 99 Mo production rate based on the Bateman equation considering 99Mo production and extraction rates. The nuclear data and neutron flux used in the analytic model is obtained from the MCNP calculation so that these two calculations are consistent from each other. For the purpose of verifying formulations and calculations schemes, the analytic calculation was conducted for a single pellet of the target assembly and the results were compared to those of independent transmutation calculations by ORIGEN2.2 code. Note that the target assembly used for the verification test is an earlier version of target assembly which has 56 target rods. This model is used for the verification test because it has diverse neutron spectra such that there are pellets of which the neutron cross sections are th~ same as those of the ORIGEN2.2 single group library. The earlier model also adopts a more complicated process during the target irradiation when compared with the once-through operation scheme of the reference target assembly. The verification test model is as follows:

92

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B 99 99

• The target assembly is continuously irradiated for Mo production and Mo is continuously extracted.

• The 99Mo is extracted three times a week and collected in the hot cell. The extraction rate is 60% in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, i.e., 2.06x10-S /sec. The collected 99 Mo is delivered every week.

• The operation continues for 1 year and the comparison is made at the end of 52nd week.

In the target assembly, each pellet has its own cross sections, while the ORIGEN2.2 uses one-group cross section libraries. For a fair comparison between the analytic model and ORIGEN2.2, a single pellet, of which the cross sections are the same as those of ORIGEN2.2, was searched. The selected pellet is pellet number 43 (from the bottom) of rod 33. The cross sections of the assembly and the selected pellet are compared with the ORIGEN2.2 library in Table 5-7.

The difference of 99 Mo content between the analytic model and ORIGEN2.2 calculation is 0.07%

and 0.2% for the target pellet and collection system, respectively. The variation of 99Mo is shown in Figure 5-14 and Figure 5-15 for the target pellet and collection system, respectively. It can be seen that the calculation results of both analytic and ORIGEN2.2 calculations are overlapping.

99 This confirms that the physics model has been correctly formulated to predict Mo content in the target assembly.

Table 5-7. Comparison of 235 U cross sections Absorption (barn) Fission (barn)

ORIGEN2.2 57.2 46.7 Rod33-Pellet43 (MCNP) 57.1 46.9 56-rod assembly-average (MCNP) 114.8 96.3 93 l

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Throug h Operation 3044 1R0003 1/B I

- Pellet43 Mo99Analytic

- Pellet43 Mo99 ORIGEN 0 30 60 90 Day 99 Figure 5-14. Comparison Mo content in the pellet between analytic model and ORIGEN2.2 T

- ORIGEN fin ite difference

- - Analytic

+-~~-l-~~-f-~~--+-~~-i-~~--+-~~--l~-----l 357 358 359 360 361 362 363 364 Day 99 Figure 5-15. Comparison Mo content in the collection system between analytic model and ORIGEN2.2 6

SUMMARY

The nuclear performance of the target assembly has been evaluated using a single target pellet enrichment of - wt% and aod loading per assembly. The evaluation results of the reference target assembly model are as follows:

94

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R0003 1/B

• The stand-alone - assemblies together have a sufficiently low sub-criticality below 0.659.

• The MURR driver fuel element power peaking factor is less than 3.27 when

• fresh target assemblies are loaded.

• The variation of target pellet linear power and total target assembly power due to CB movement is small . For the equilibrium core conditions, the peak linear power and total target assembly power change from - to - kW/m ~%)and from - to

- kW . .%), respectively, when the control blades travel from position 46 cm to 62cm.

• The variation of the eigenvalue due to pellet manufacturing tolerance does not exceed

- when the fabrication density changes by - or the enrichment changes by The uncertainty due to mechanical tolerance of the target rod loading -

confidence level) is in 95%

for the peak linear power of the extreme burnup core and total target power of the maximum burnup core, respectively.

• The peak linear power and total target power of the staggered and partial loading are less than those of the base loading.

Nuclear compatibility of the target assembly with MURR core has been assessed based on applicable MURR technical specifications. In general, because the power from the target assembly is relatively small e 10 of the nominal MURR core power for the full target assembly loading) and the target assemblies are loaded outside the beryllium reflector, the effect of target assembly on the core reactivity characteristics is small as summarized below for representative core burnup states with the base target assembly loading:

• The least negative core primary coolant temperature coefficient of reactivity is -

-

• which is more negative than -1 .08><104 (.D.k/k)/°C of Technical Specification 5.3.a. The primary coolant temperature range in this calculation was from 20.44°C to 76.84°C.

• The core primary coolant void coefficient of reactivity is , which is more negative than -2.0><10*3 (.D.k/k)/%void of Technical Specification 5.3.b. The primary coolant density in the calculation was changed from 100% nominal to 0.1%.

• The highest regulating rod reactivity worth is which is less than 6.0><10"3

.D.k/k of Technical Specification 5.3.d.

95

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B

• The lowest subcriticality margin of the core with the most reactive shim blade and regulating blade fully withdrawn i s -

• which is greater than 0.02 Lik/k of the Technical Specification 3.1.b.

• The excess reactivity of the cold, clean, minimum burnup core with cold, fresh, base target assembly loading i s -

• which is less than 0.098 Lik/k referenced to the cold, clean, critical core in Technical Specification 3.1.a.

• A single target assembly insertion into the MURR pool adds - to the MURR core, which is lower than allowed reactivity worth of each secured removable experiment, 0.6%.Lik (Technical Specification: 3.8.a).

7 REFERENCES

[ASME 2011] Materials. Part D, Properties (Metric), New York, American Society of Mechanical Engineers, 2011.

[ATI 2016] Reactor Grade Zirconium Alloys for Nuclear Waste Disposal, https://www.atimetals.com, 2016.

[Bateman 191 O] Bateman, H., "Solution of a System of Differential Equations Occurring in the Theory of Radioactive Transformation," Proc. Cambridge Phil. Soc.,

1910.

[Bolin 2016] Bolin, J., Private Communication, General Atomics, May 2016.

[Briesmeister, 2000] Briesmeister, J. F., Ed., "MCNP' - A General Monte Carlo N-Particle Transport Code," LA-13709-M, Los Alamos National Laboratory, March 2000.

[Choi 2015b] Choi, H., "Verification of Physics Design Model for Mass Production of 99 Mo in a Nuclear Reactor," ASME 2015 Verification and Validation Symposium, Las Vegas, May 2015.

96

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B

[Conlin 2013] Conlin, J. L. et al., "Continuous Energy Neutron Cross Section Data Tables Based upon ENDF/B-Vll.1," LA-UR-13-20137, Los Alamos National Laboratory, Feb. 2013.

[Croff 1980] Croff, A. G., "A User's Manual for the ORIGEN2 Computer Code,"

ORNL/TM-7175, Oak Ridge national Laboratory, July 1980.

[Deen 2003] Deen, J. R., et al., "WIMS-ANL User Manual, Rev. 5," ANL/TD/TM99-07, Argonne National Laboratory, 2003.

[Dionne 2008] Dionne, B., Stillman, J., Stevens, J., "Applicability of WIMS-ANL to Generate Cross Sections for Very High Density UMo Fuel in Proposed LEU Fuel Assembly," Proc. 2008 International Meeting on Reduced Enrichment for Research and Test Reactors, Washington, D.C., Oct. 5-9, 2008.

[Durkee 2012] Durkee Jr., J. W., et al., "The MCNP6 Delayed-Particle Feature," LA-UR-12-00283, Los Alamos National Laboratory, March 2012.

[Foyto 2014] Foyto, L. et al., "Accident Analyses for the Conversion of the University of Missouri Research Reactor from Highly-Enriched to Low-Enriched Uranium," 35th Int. Mtg. on Reduced Enrichment for Research and Test Reactors, RERTR 2014, Vienna, Austria, Oct. 2014.

[Goorley 2013a] Goorley, J. T. et al., "Initial MCNP6 Release Overview - MCNP6 version 1.0," LA-UR-13-:22934, Los Alamos National Laboratory, April 2013.

[Goorley 2013b] Goorley, J. T. et al., "Features of MCNP6," Joint International Conference on Supercomputing in Nuclear Applications and Monte Carlo 2013 (SNA

+ MC2013), Paris, France, October 27-31, 2013.

[Hanan 1998] Hanan, N. A. et al., "The Use of WIMS-ANL Lumped Fission Product Cross Sections for Burned Core Analysis with the MCNP Monte Carlo Code," Proc. 1998 International Meeting on Reduced Enrichment for Research and Test Reactors, Sao Paulo, Brazil, Oct. 18-23, 1998.

[Kiedrowski 201 O] Kiedrowski, B. C. et al., "MCNPS-1.60 Feature Enhancement & Manual Clarifications," LA-UR-10-06217, Los Alamos national Laboratory, 2010.

97

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00031/B

[Kim 2005] Kim, S. S. et al., "Advanced Test Reactor: Serpentine Arrangement of Highly Enriched Water-Moderated Uranium-Aluminide Fuel Plates Reflected by Beryllium," NEA/NSC/DOC/(95)03/11, Volume II, Nuclear Energy Agency, September 2005.

[LANL 2003] * "MCNP - A General Monte Carlo N-Particle Transport Core, Version 5,"

LA-UR-03-1987, Los Alamos National Laboratory, April 2003.

[Lawrence 1983] Lawrence, R. D., "The DIF3D Nodal Neutronics Option for Two- and Three-Dimensional Diffusion-Theory Calculations in Hexagonal Geometry," ANL-83-1, Argonne National Laboratory, March 1983.

[Marshall 2013] Marshall, F. M., "The Advanced Test Reactor National Scientific User Facility Overview," IAEA, June 2013.

[McKibben 2006] McKibben, J. C. et al., "Current Status of the Missouri University Research Reactor HEU to LEU Conversion Feasibility Study," 2006 Int.

RERTR Meeting, Cape Town, South Africa, Oct. 29- Nov. 2, 2006.

[McKibben 2009] McKibben, J. C., et al., "Feasibility Analyses for HEU to LEU Fuel Conversion of the University of Missouri Research Reactor (MURR),"

MURR Technical Data Report No. 0125, September 2009.

[Mosteller 2002] Mosteller, R. D., Validation Suites for MCNP," LA-UR-02-0878, Los Alamos National Laboratory, 2002.

[Mosteller 2010] Mosteller, R. D., "An Expanded Criticality Validation Suites for MCNP,"

LA-UR-10-06230, Los Alamos National Laboratory, 2010.

[Motojima 1977] Motojima, K. et al., "Volatilization Process for Separation of Molybdenum-99 from Irradiated Uranium," United States Patent 4,017;583, April 12, 1~77.

[MURR2006] "Missouri University Research Reactor Technical Specifications," Docket 50-186, License R-103, University of Missouri, Columbia, August 2006.

[Nakagawa 2003] Nakagawa, T., Ed., "The JENDL-3.3 Fission Product Yield Data Library,"

IAEA-NDS-138, International Atomic Energy Agency, January 2003.

[NEA2010] "International Handbook of Evaluated Reactor Physics Benchmark Experiments," NEA/NSC/DOC/(2006)1, Nuclear Energy Agency, March 2010.

98

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B

[NRC 1997] "Standard review plan for dry cask storage systems," U.S. Nuclear Regulatory Commission, Jan. 1997.

[NRC 1999] "Safety evaluation by the office of nuclear reactor regulation supporting amendment No. 31 to amended facility operating license No. R-37,"

Docket No. 50-20, Nuclear Regulatory Commission, Dec. 1999.

[Olson 2001] Olson, A. P., A User's Guide for the REBUS-PC Code, Version 1.4, ANURERTR/TM-32, Argonne National Laboratory, 2001

[Parker 2015] Parker, E. M. et al., "Y-12 Standard Specification Low Enriched Uranium Metal Supply for 99Mo Isotope Production," Y/PM-462, National Nuclear Security Administration, August 2015.

[Parsons 2012] Parsons, D. K., Conlin, J. L, "Release of Continuous Representation for S(a,13) ACE Data," LA-UR-12-00800, Los Alamos National Laboratory, Feb. 2012.

[Pelowitz 2013] Pelowitz, D. B., Ed., "MCNP6' User's Manual Version 1.0," LA-CP 00634, Los Alamos National Laboratory, May 2013.

[Peters 2012a] Peters, N. J., Kutikkad, K., "Improvements to Predictions of the Estimated Critical Blade Positions Due to Control Blade Depletion Corrections at MURR," 34th Int. Mtg. on Reduced Enrichment for Research and Test Reactors, RERTR 2012, Warsaw, Poland, Oct. 2012.

[Peters 2012b] Peters, N., Kutikkad, K., "MURR Control Blade Depletion Study Report,"

MURR Technical Data Report No. 134, June 2012.

[Peters 2013] Peters, N. J., Kutikkad, K., "Validation for Computational Methods and Calculations at MURR (Draft)," University of Mi.ssouri Research Reactor, December 2013.

[Rosenbaum 1978] Rosenbaum, H. S. et al., "Process for Separating Fission Product Molybdenum from an Irradiated Target Material," United States Patent 4,123,498, Oct. 31, 1978.

[Stillman 2012] Stillman, J. A. et al., Technical Basis in Support of the Conversion of the University of Missouri Research Reactor (MURR) Core from Highly-99

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Enriched to Low-Enriched Uranium - Core Neutron Physics,"

ANURERTR!TM-12-30, Argonne National laboratory, September 2012.

[Stillman 2013] Stillman, J. A., "Conceptual Design Parameters for MURR LEU U-Mo Fuel Conversion Design Demonstration Experiment," ANURERTR!TM-12-28 Revision 1, Argonne National laboratory, March 2013.

[Trellue 1999] Trellue, H. R., Poston, D. I., "User's Manual, Version 2.0 for Monteburns, Version SB," LA-UR-99-4999, Los Alamos National Laboratory, 1999.

[Whalen 1992] Whalen, D. J. et al., "MCNP: Neutron Benchmark Problems," LA-12212, Los Alamos National Laboratory, 1992.

[Whipple 2009] Whipple, C., et al., "Medical Isotope Production without Highly Enriched Uranium," The National Academies Press, Washington, D.C., 2009.

[White 2000] White, M. C., "Development and Implementation of Photonuclear Cross-

• Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code," LA-13744-T, Los Alamos National Laboratory, July 2000.

[White 2003] White, M. C., "Photoatomic Data Library MCPLIB04: A New Photoatomic Library Based On Data from ENDF/B-VI Release 8," LA-UR-03-1019, Los Alamos National Laboratory, February 2003.

[Xu 2006] Xu, Z.,* et al, "MCODE, Version 2.2 - An MCNP-ORIGEN Depletion Program," Center for Advanced Nuclear Energy Systems (CANES),

Nuclear Science & Engineering Department, Massachusetts Institute of Technology, Cambridge, MA, February 2006.

100

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B APPENDIX A MURR FUEL ELEMENT PEAKING FACTORS A-1

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Table A-1. MURR fuel element 1 peaking factor for the maximum burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7. 8 9 10 11 12 24 0.940 0.771 0.664 0.610 0.551 0.516 0.487 0.469 0.448 0.427 0.410 0.401 23 0.963 0.747 0.632 0.560 0.505 0.461 0.423 0.403 0.378 0.363 0.352 0.338 22 1.110 0.870 0.734 0.649 0.579 0.536 0.488 0.470 0.446 0.422 0.409 0.399 21 1.275 1.008 0.844 0.750 0.676 .0.627 0.580 0.551 0.522 0.507 0.482 0.462 20 1.438 1.153 0.960 0.861 0.765 0.702 0.653 0.625 0.595 0.571 0.549 0.535 19 1.610 1.284 1.077 0.952 0.858 0.796 0.746 0.701 0.670 0.647 0.619 0.613 18 1.790 1.419 1.197 1.059 0.964 0.893 0.835 0.790 0.749 0.721 0.699 0.678 17 1.985 1.580 1.344 1.188 1.074 0.999 0.929 0.878 0.847 0.808 0.788 0.765 16 2.152 1.722 1.462 1.305 1.178 1.085 1.025 0.975 0.936 0.893 0.874 0.854 15 2.319 1.864 1.587 1.392 1.280 1.182 1.108 1.052 1.006 0.973 0.948 0.940 14 2.473 1.981 1.675 1.472 1.340 1.257 1.190 1.130 1.084 1.062 1.019 1.010 13 2.558 2.050 1.742 1.564 1.412 1.316 1.238 1.198 1.139 1.115 1.096 1.071 12 2.592 2.102 1.791 1.593 1.453 1.354 1.281 1.229 1.183 1.155 1.129 1.110 11 2.636 2.122 1.813 1.611 1.478 1.386 1.301 1.249 1.211 1.184 1.148 1.143 10 2.652 2.150 1.817 1.625 1.496 1.379 1.313 1.261 1.214 1.197 1.165 1.154 9 2.704 2.165 1.839 1.628 1.492 1.389 1.311 1.255 1.217 1.178 1.163 1.141 8 2.663 2.129 1.809 1.599 1.461 1.350 1.269 1.230 1.186 1.152 1.133 1.116 7 2.564 2.032 1.734 1.555 1.419 1.316 1.253 1.197 1.148 1.124 1.098 1.078 6 2.448 1.947 1.673 1.480 1.351 1.246 1.178 1.132 1.091 1.068 1.038 1.020 5 2.283 1.826 1.550 1.371 1.255 1.157 1.101 1.057 1.032 1.000 0.979 0.959 4 2.061 1.653 1.418 1.255 1.147 1.067 1.013 0.959 0.933 0.906 0.894 0.885 3 1.878 1.498 1.280 1.135 1.034 0.963 0.907 0.862 0.838 0.821 0.796 0.788 2 1.679 1.342 1.148 1.013 0.936 0.852 0.811 0.770 0.750 0.730 0.716 0.705 1 1.715 1.430 1.274 1.154 1.075 1.019 0.979 0.943 0.915 0.897 0.887 0.880 Plate 2.046 1.640 1.396 1.240 1.130 1.049 0.989 0.945 0.909 0.883 0.861 0.846 Axial 1.338 1.338 1.335 1.330 1.340 1.342 1.345 1.352 1.356 1.373 1.371 1.381 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.388 0.374 0.364 0.352 0.339 0.329 0.318 0.303 0.293 0.288 0.278 0.279 23 0.328 0.320 0.312 0.306 0.301 0.292 0.283 0.278 0.270 0.265 0.265 0.275 22 0.389 0.380 0.367 0.355 0.347 0.334 0.328 0.319 0.317 0.313 0.309 0.313 21 0.448 0.438 0.429 0.416 0.401 0.393 0.376 0.372 0.370 0.356 0.361 0.374 20 0.519 0.503 0.492 0.473 0.460 0.453 0.441 0.436 0.423 0.418 0.421 0.427 19 0.596 0.574 0.565 0.550 0.534 0.520 0.516 0.501 0.491 0.479 0.479 0.485 18 0.659 0.639 0.630 0.613 0.605 0.578 0.568 0.560 0.555 0.549 0.546 0.564 17 0.746 0.725 0.711 0.694 0.682 0.665 0.650 0.642 0.629 0.628 0.639 0.667 16 0.838 0.819 0.800 0.783 0.777 0.767 0.759 0.761 0.766 0.788 0.823 0.904 15 0.920 0.903 0.896 0.892 0.885 0.898 0.910 0.934 0.973 1.049 1.151 1.351 14 0.991 0.985 0.979 0.981 0.993 1.002 1.030 1.080 1.146 1.250 1.406 1.696 13 1.062 1.050 1.052 1.055 1.058 1.079 1.102 1.155 1.242 1.371 1.567 1.899 12 1.093 1.092 1.091 1.088 1.103 1.132 1.168 1.217 1.299 1.433 1.647 2.010 11 1.133 1.111 1.114 1.118 1.129 1.154 1.197 1.260 1.349 1.475 1.694 2.072 10 1.140 1.124 1.118 1.134 1.150 1.175 1.211 1.266 1.358 1.508 1.716 2.091 9 1.133 1.118 1.127 1.135 1.150 1.166 1.212 1.265 1.364 1.499 1.711 2.101 8 1.109 1.102 1.107 1.118 1.124 1.148 1.189 1.252 1.336 1.470 1.689 2.072 7 1.067 1.066 1.071 1.075 1.099 1.097 1.151 1.213 1.292 1.433 1.642 2.006 6 1.012 1.014 1.014 1.022 1.044 1.054 1.086 1.142 1.229 1.369 1.560 1.912 5 0.960 0.952 0.954 0.957 0.975 0.989 1.025 1.072 1.153 1.277 1.467 1.791 4 0.872 0.862 0.861 0.859 0.875 0.902 0.936 0.989 1.065 1.180 1.362 1.658 3 0.775 0.764 0.777 0.778 0.796 0.815 0.844 0.886 0.960 1.060 1.227 1.514 2 0.697 0.694 0.696 0.705 0.725 0.726 0.755 0.800 0.856 0.954 1.113 1.378 1 0.872 0.868 0.857 0.872 0.872 0.871 0.891 0.922 0.975 1.052 1.181 1.412 Plate 0.834 0.822 0.818 0.816 0.820 0.825 0.842 0.870 0.916 0.990 1.108 1.319 Axial 1.386 1.385 1.396 1.409 1.421 1.443 1.459 1.473 1.508 1.542 1.569 1.613 A-2

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-2. MURR fuel element 2 peaking factor for the maximum burnup core - )

Axial Fuel plate number node 2 3 4 5 6 7 8 9 10 11 12 24 0.790 0.683 0.625 0.552 0.518 0.487 0.467 0.447 0.423 0.408 0.398 0.382 23 0.811 0.678 0.588 0.525 0.475 0.435 0.412 0.391 0.377 0.351 0.343 0.334 22 0.917 0.762 0.671 0.594 0.539 0.507 0.472 0.444 0.421 0.410 0.398 0.383 21 1.048 0.888 0.776 0.703 0.632 0.589 0.556 0.521 0.507 0.489 0.470 0.451 20 1.117 0.973 0.869 0.782 0.724 0.668 0.625 0.600 0.572 0.548 0.535 0.514 19 1.250 1.093 0.983 0.886 0.812 0.754 0.705 0.674 0.649 0.623 0.605 0.586 18 1.333 1.182 1.046 0.958 0.885 0.827 0.774 0.749 0.721 0.681 0.667 0.646 17 1.477 1.303 1.167 1.064 0.987 0.923 0.859 0.829 0.788 0.759 0.740 0.719 16 1.571 1.400 1.263 1.153 1.073 0.996 0.942 0.900 0.862 0.846 0.819 0.800 15 1.700 1.519 1.380 1.241 1.151 1.080 1.019 0.987 0.953 0.929 0.898 0.876 14 1.755 1.570 1.420 1.308 1.213 1.139 1.087 1.040 0.999 0.973 0.958 0.943 13 1.833 1.640 1.505 1.379 1.276 1.193 1.138 1.093 1.055 1.038 1.015 0.998 12 1.849 1.689 1.544 1.412 1.320 1.232 1.169 1.114 1.084 1.076 1.043 1.035 11 1.893 1.695 1.542 1.414 1.314 1.252 1.193 1.149 1.113 1.096 1.076 1.065 10 1.916 1.708 1.564 1.436 1.338 1.262 1.195 1.156 1.119 1.103 1.094 1.065 9 1.905 . 1.711 1.553 1.434 1.329 1.243 1.181 1.148 1.124 1.100 1.071 1.057 8 1.913 1.701 1.539 1.403 1.316 1.234 1.183 1.136 1.104 1.082 1.056 1.048 7 1.853 1.658 1.508 1.367 1.279 1.204 1.141 1.093 1.050 1.037 1.020 1.011 6 1.826 1.612 1.448 1.322 1.224 1.152 1.097 1.054 1.020 0.991 0.979 0.960 5 1.725 1.509 1.360 1.239 1.142 1.081 1.027 0.979 0.952 0.926 0.916 0.903 4 1.631 1.431 1.271 1.150 1.064 0.994 0.946 0.912 0.883 0.860 0.844 0.834 3 1.456 1.264 1.131 1.020 0.940 0.879 0.836 0.796 0.783 0.766 0.747 0.737 2 1.359 1.164 1.021 0.929 0.854 0.799 0.761 0.730 0.704 0.690 0.676 0.668 1 1.381 1.229 1.120 1.050 0.994 0.947 0.923 0.885 0.857 0.846 0.841 0.819 Plate 1.698 1.499 1.351 1.231 1.141 1.070 1.015 0.974 0.941 0.918 0.898 0.881 Axial 1.267 1.281 1.299 1.309 1.316 1.324 1.321 1.332 1.341 1.349 1.367 1.357 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.372 0.358 0.350 0.339 0.322 0.303 0.291 0.277 0.261 0.244 0.234 0.222 23 0.324 0.312 0.298 0.290 0.284 0.277 0.268 0.252 0.240 0.231 0.222 0.212 22 0.370 0.360 0.350 0.342 0.329 0.315 0.303 0.290 0.283 0.264 0.251 0.236 21 0.438 0.432 0.419 0.404 0.392 0.376 0.361 0.346 0.331 0.314 0.295 0.275 20 0.503 0.480 0.469 0.453 0.437 0.426 0.402 0.387 0.367 0.344 0.320 0.299 19 0.563 0.550 0.539 0.515 0.498 0.485 0.462 0.447 0.424 0.402 0.377 0.346 18 0.619 0.609 0.589 0.571 0.555 0.536 0.518 0.492 0.470 0.437 0.407 0.374 17 0.695 0.683 0.659 0.651 0.632 0.608 0.589 0.572 0.545 0.522 0.478 0.455 16 0.778 0.758 0.747 0.732 0.729 0.711 0.701 0.688 0.673 0.663 0.653 0.636 15 0.864 0.851 0.848 0.844 0.847 0.848 0.856 0.873 0.903 0.940 0.999 1.078 14 0.936 0.916 0.920 0.930 0.939 0.956 0.981 1.012 1.067 1.142 1.235 1.362 13 0.989 0.984 0.982 0.989 1.004 1.025 1.064 1.100 1.169 1.249 1.368 1.515 12 1.022 1.011 1.022 1.030 1.049 1.077 1.111 1.157 1.235 1.327 1.443 1.594 11 1.045 1.050 1.045 1.052 1.079 1.105 1.135 1.190 1.261 1.354 1.473 1.631 10 1.057 1.056 1.056 1.064 1.088 1.114 1.157 1.208 1.279 1.379 1.504 1.666 9 1.049 1.046 1.051 1.059 1.075 1.106 1.138 1.196 1.272 1.363 1.498 1.664 8 1.032 1.028 1.032 1.039 1.062 1.082 1.131 1.181 1.261 1.358 1.503 1.666 7 0.994 0.998 0.997 1.004 1.021 1.048 1.094 1.144 1.207 1.319 1.441 1.608 6 0.959 0.954 0.955 0.975 0.988 1.003 1.039 1.096 1.174 1.270 1.409 1.595 5 0.898 0.896 0.886 0.898 0.921 0.948 0.973 1.018 1.084 1.188 1.316 1.490 4 0.831 0.829 0.825 0.838 0.862 0.868 0.907 . 0.951 1.024 1.126 1.244 1.422 3 0.730 0.733 0.739 0.746 0.751 0.776 0.815 0.849 0.914 1.000 1.113 1.282 2 0.669 0.663 0.660 0.666 0.685 0.701 0.735 0.767 0.825 0.913 1.031 1.200 1 0.825 0.816 0.816 0.821 0.830 0.837 0.859 0.889 0.935 0.993 1.085 1.227 Plate 0.868 0.859 0.854 0.853 0.859 0.867 0.883 0.906 0.945 0.998 1.071 1.172 Axial 1.366 1.380 1.388 1.399 1.421 1.443 1.470 1.496 1.519 1.551 1.577 1.596 A-3

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-3. MURR fuel element 3 peaking factor for the maximum burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.901 0.760 0.687 0.619 0.578 0.546 0.518 0.496 0.472 0.459 0.449 0.436 23 0.933 0.767 0.655 0.576 0.524 0.485 0.452 0.429 0.412 0.400 0.383 0.376 22 1.053 0.863 0.746 0.664 0.610 0.560 0.528 0.508 0.485 0.473 0.454 0.442 21 1.200 0.982 0.861 0.776 0.699 0.651 0.622 0.595 0.568 0.545 0.536 0.521 20 1.308 1.101 0.964 0.869 0.799 0.745 0.705 0.682 0.652 0.630 0.610 0.600 19 1.439 1.214 1.066 0.975 0.898 0.848 0.799 0.756 0.738 0.717 0.707 0.689 18 1.567 1.334 1.181 1.081 0.998 0.923 0.886 0.848 0.817 0.794 0.773 0.764 17 1.761 1.499 1.319 1.200 1.107 1.042 0.982 0.940 0.912 0.886 0.875 0.858 16 1.885 1.624 1.429 1.297 1.203 1.128 1.052 1.006 0.990 0.950 0.941 0.924 15 2.050 1.755 1.554 1.389 1.281 1.205 1.138 1.080 1.046 1.030 1.013 0.998 14 2.136 1.834 1.620 1.453 1.343 1.273 1.194 1.151 1.115 1.087 1.063 1.052 13 2.241 1.925 1.690 1.537 1.414 1.314 1.233 1.192 1.166 1.132 1.109 1.089 12 2.255 1.946 1.713 1.563 1.435 1.356 1.276 1.229 1.195 1.158 1.135 1.116 11 2.306 1.964 1.741 1.572 1.459 1.366 1.301 1.256 1.207 1.177 1.141 1.121 10 2.322 1.981 1.751 1.584 1.462 1.369 1.316 1.265 1.216 1.178 1.155 1.140 9 2.288 1.971 1.728 1.563 1.442 1.358 1.285 1.238 1.197 1.176 1.147 1.125 8 2.310 1.947 1.717 1.549 1.423 1.334 1.272 1.216 1.171 1.146 1.127 1.108 7 2.216 1.896 1.666 1.490 1.376 1.287 1.214 1.165 1.121 1.101 1.079 1.060 6 2.144 1.806 1.575 1.425 1.317 1.228 1.156 1.105 1.075 1.050 1.023 1.012 5 1.968 1.670 1.454 1.316 1.207 1.140 1.064 1.033 0.986 0.968 0.947 0.933 4 1.820 1.540 1.347 1.225 1.119 1.036 0.985 0.955 0.920 0.893 0.867 0.853 3 1.653 1.372 1.205 1.081 0.992 0.929 0.879 0.838 0.811 0.790 0.781 0.776 2 1.510 1.248 1.081 0.970 0.890 0.820 0.785 0.749 0.728 0.712 0.697 0.687 1 1.563 1.340 1.207 1.088 1.035 0.981 0.950 0.930 0.899 0.886 0.868 0.861 Plate 1.767 1.499 1.318 1.190 1.098 1.028 0.973 0.935 0.903 0.880 0.861 0.847 Axial 1.301 1.309 1.315 1.317 1.319 1.318 1.338 1.340 1.333 1.325 1.327 1.332 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.421 0.412 0.402 0.387 0.375 0.365 0.353 0.344 0.333 0.325 0.319 0.326 23 0.369 0.358 0.349 0.340 0.334 0.332 0.326 0.319 0.317 0.320 0.327 0.339 22 0.436 0.419 0.414 0.403 0.401 0.391 0.389 0.377 0.375 0.373 0.385 0.404 21 0.512 0.502 0.498 0.486 0.481 0.476 0.478 0.479 0.477 0.485 0.501 0.538 20 0.580 0.582 0.574 0.568 0.562 0.573 0.575 0.586 0.597 0.623 0.669 0.738 19 0.686 0.676 0.663 0.665 0.666 0.666 0.675 0.693 0.710 0.757 0.834 0.945 18 0.756 0.750 0.745 0.736 0.742 0.751 0.760 0.775 0.807 0.864 0.935 1.055 17 0.835 0.823 0.823 0.818 0.826 0.831 0.847 0.880 0.917 0.972 1.068 1.203 16 0.906 0.898 0.898 0.895 0.911 0.911 0.935 0.970 1.023 1.093 1.203 1.361 15 0.991 0.979 0.978 0.973 0.991 1.015 1.049 1.078 1.151 1.250 1.391 1.611 14 1.037 1.047 1.033 1.037 1.060 1.075 1.116 1.175 1.250 1.356 1.523 1.754 13 1.089 1.072 1.083 1.087 1.107 1.127 1.174 1.227 1.316 1.433 1.605 1.869 12 1.112 1.098 1.101 1.119 1.138 1.175 1.207 1.263 1.349 1.472 1.651 1.916 11 1.108 1.111 1.112 1.127 1.145 1.174 1.221 1.283 1.358 1.481 1.669 1.944 10 1.128 1.128 1.122 1.139 1.160 1.193 1.228 1.287 1.369 1.505 1.683 1.943 9 1.117 1.111 1.113 1.125 1.145 1.164 1.209 1.271 1.365 1.473 1.643 1.914 8 1.095 1.090 1.091 1.090 1.105 1.132 1.173 1.234 1.318 1.435 1.604 1.873 7 1.044 1.040 1.041 1.044 1.066 1.096 1.126 1.186 1.264 1.388 1.553 1.802 6 1.003 0.997 0.993 1.005 1.017 1.048 1.082 1.131 1.199 1.326 1.484 1.725 5 0.934 0.923 0.918 0.929 0.941 0.961 0.996 1.041 1.113 1.223 1.375 1.612 4 0.851 0.844 0.851 0.861 0.870 0.886 0.916 0.968 1.026 1.123 1.277 1.507 3 0.761 0.759 0.760 0.761 0.768 0.786 0.815 0.861 0.933 1.020 1.157 1.377 2 0.678 0.673 0.674 0.684 0.707 0.719 0.736 0.777 0.839 0.934 1.064 1.275 1 0.846 0.838 0.840 0.825 0.849 0.859 0.875 0.903 0.955 1.024 1.130 1.306 Plate 0.837 0.830 0.828 0.829 0.840 0.854 0.877 0.912 0.964 1.042 1.157 1.334 Axial 1.334 1.344 1.342 1.360 1.367 1.383 1.387 1.397 1.406 1.430 1.440 1.443 A-4

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-4. MURR fuel element 4 peaking factor for the maximum burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.834 0.734 0.676 0.622 0.590 0.550 0.534 0.512 0.496 0.487 0.471 0.458 23 0.846 0.724 0.647 0.580 0.541 0.501 0.477 0.460 0.444 0.427 0.412 0.402 22 0.955 0.835 0.738 0.677 0.623 0.574 0.551 0.522 0.508 0.493 0.486 0.471 21 1.085 0.961 0.851 0.778 0.732 0.678 0.647 0.631 0.602 0.585 0.567 0.562 20 1.148 1.018 0.928 0.854 0.806 0.762 0.737 0.704 0.684 0.671 0.654 0.645 19 1.260 1.137 1.045 0.970 0.914 0.863 0.825 0.802 0.785 0.774 0.756 0.742 18 1.335 1.207 1.122 1.053 0.992 0.939 0.902 0.867 0.849 0.841 0.822 0.816 17 1.470 1.356 1.244 1.162 1.089 1.034 0.997 0.962 0.929 0.920 0.908 0.900 16 1.563 1.437 1.318 1.232 1.165 1.112 1.064 1.015 0.994 0.970 0.960 0.959 15 1.654 1.534 1.413 1.309 1.242 1.174 1.126 1.093 1.075 1.046 1.025 1.018 14 1.701 1.571 1.463 1.359 1.272 1.207 1.164 1.126 1.115 1.095 1.066 1.059 13 1.762 1.638 1.535 1.416 1.341 1.262 1.206 1.171 1.148 1.119 1.095 1.094 12 1.801 1.665 1.541 1.444 1.361 1.287 1.247 1.200 1.163 1.142 1.130 1.120 11 1.821 1.684 1.558 1.463 1.367 1.296 1.243 1.203 1.179 1.150 1.138 1.127 10 1.835 1.682 1.566 1.466 1.375 1.306 1.265 1.209 1.180 1.157 1.143 1.119 9 1.812 1.686 1.561 1.436 1.360 1.293 1.238 1.190 1.161 1.137 1.117 1.101 8 1.837 1.682 1.547 1.441 1.354 1.271 1.228 1.179 1.153 1.122 1.099 1.084 7 1.771 1.615 1.492 1.382 1.293 1.228 1.164 1.118 1.104 1.075 1.057 1.051 6 1.769 1.598 1.445 1.350 1.244 1.172 1.120 1.080 1.046 1.034 1.017 1.004 5 1.641 1.469 1.336 1.231 1.147 1.080 1.041 1.001 0.975 0.943 0.942 0.929 4 1.560 1.385 1.253 1.145 1.061 0.996 0.957 0.931 0.904 0.877 0.868 0.855 3 1.410 1.252 1.129 1.036 0.966 0.908 0.870 0.828 0.805 0.779 0.772 0.760 2 1.305 1.137 1.021 0.941 0.873 0.808 0.778 0.740 0.721 0.710 0.692 0.687 1 1.334 1.209 1.122 1.042 0.989 0.952 0.910 0.892 0.866 0.857 0.847 0.834 Plate 1.473 1.336 1.226 1.136 1.066 1.006 0.966 0.931 0.908 0.888 0.873 0.863 Axial 1.241 1.256 1.272 1.284 1.284 1.292 1.303 1.293 1.294 1.297 1.303 1.301 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.447 0.430 0.424 0.412 0.400 0.383 0.374 0.357 0.336 0.318 0.306 0.290 23 0.393 0.382 0.377 0.365 0.357 0.351 0.343 0.333 0.319 0.309 0.304 0.298 22 0.454 0.453 0.439 . 0.435 0.423 0.421 0.407 0.399 0.383 0.370 0.362 0.353 21 0.547 0.545 0.537 0.540 0.534 0.520 0.520 0.516 0.517 0.523 0.526 0.546 20 0.638 0.637 0.638 0.634 0.641 0.639 0.652 0.669 0.692 0.730 0.773 0.844 19 0.738 0.739 0.742 0.756 0.753 0.773 0.791 0.823 0.861 0.923 1.001 1.114 18 0.804 0.806 0.815 0.824 0.831 0.851 0.884 0.921 0.961 1.024 1.101 1.215 17 0.891 0.894 0.903 0.911 0.923 0.944 0.978 1.016 1.071 1.153 1.248 1.367 16 0.952 0.943 0.952 0.968 0.976 1.008 1.040 1.085 1.150 1.239 1.334 1.462 15 1.000 1.008 1.019 1.029 1.044 1.077 1.117 1.169 1.245 1.339 1.446 1.592 14 1.045 1.041 1.050 1.079 1.097 1.120 1.158 1.219 1.300 1.392 1.501 1.639 13 1.087 1.089 1.104 1.116 1.140 1.167 1.216 1.272 1.345 1.449 1.561 1.702 12 1.110 1.106 1.111 1.125 1.147 1.179 1.226 1.282 1.364 1.462 1.574 1.722 11 1.118 1.126 1.125 1.149 1.170 1.196 1.237 1.296 1.381 1.476 1.596 1.727 10 1.110 1.120 1.141 1.140 1.173 1.202 1.245 1.300 1.373 1.479 1.598 1.742 9 1.096 1.100 1.104 1.118 1.148 1.181 1.227 1.277 1.360 1.456 1.573 1.716 8 1.075 1.078 1.089 1.102 1.121 1.158 1.199 1.261 1.337 1.437 1.557 1.711 7 1.038 1.031 1.036 1.052 1.067 1.103 1.147 1.198 1.279 1.377 1.494 1.647 6 0.996 0.999 1.003 1.008 1.033 1.057 1.098 1.151 1.225 1.322 1.447 1.613 5 0.916 0.916 0.918 0.931 0.950 0.981 1.021 1.068 1.139 1.227 1.338 1.497 4 0.847 0.852 0.854 0.861 0.876 0.905 0.942 0.988 1.059 1.145 1.267 1.430 3 0.756 0.762 0.763 0.766 0.783 0.803 0.839 0.883 0.942 1.023 1.137 1.286 2 0.679 0.678 0.676 0.683 0.698 0.712 0.751 0.790 0.845 0.926 1.040 1.183 1 0.835 0.825 0.830 0.844 0.837 0.851 0.873 0.897 0.944 1.002 1.081 1.189 Plate 0.854 0.853 0.857 0.865 0.876 0.895 0.924 0.961 1.013 1.083 1.168 1.281 Axial 1.305 1.314 1.326 1.323 1.332 1.336 1.341 1.347 1.357 1.360 1.362 1.354 A-5

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Table A-5. MURR fuel element 5 peaking factor for the maximum burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 1.064 0.885 0.787 0.712 0.659 0.627 0.598 0.578 0.570 0.557 0.544 0.528 23 1.097 0.875 0.747 0.658 0.598 0.564 0.537 0.509 0.491 0.477 0.464 0.457 22 1.260 1.010 0.859 0.768 0.698 0.653 0.620 0.595 0.572 0.561 0.539 0.530 21 1.431 1.152 0.992 0.880 0.815 0.760 0.729 0.689 0.674 0.661 0.645 0.639 20 1.596 1.312 1.130 1.006 0.927 0.866 0.817 0.797 0.778 0.758 0.737 0.736 19 1.781 1.453 1.245 1.113 1.029 0.977 0.927 0.894 0.876 0.860 0.852 0.840 18 1.987 1.626 1.399 1.255 1.154 1.083 1.035 1.000 0.965 0.942 0.934 0.928 17 2.182 1.784 1.536 1.366 1.262 1.183 1.129 1.093 1.053 1.029 1.027 1.016 16 2.364 1.935 1.668 1.491 1.368 1.274 1.208 1.172 1.135 1.102 1.091 1.080 15 2.538 2.064 1.770 1.583 1.451 1.357 1.303 1.243 1.219 1.194 1.165 1.144 14 2.691 2.165 1.862 1.666 1.535 1.433 1.360 1.299 1.265 1.229 1.214 1.199 13 2.766 2.245 1.926 1.705 1.564 1.467 1.403 1.349 1.310 1.276 1.254 1.230 12 2.824 2.266 1.946 1.741 1.607 1.501 1.423 1.369 1.335 1.292 1.289 1.260 11 2.844 2.305 1.978 1.761 1.618 1.508 1.433 1.375 1.355 1.315 1.301 1.285 .

10 2.861 2.299 1.962 1.759 1.613 1.513 1.443 1.385 1.344 1.314 1.282 1.266 9 2.845 2.294 1.951 1.744 1.589 1.497 1.423 1.369 1.318 1.281 1.272 1.249 8 2.785 2.241 1.912 1.703 1.562 1.460 1.372 1.325 1.280 1.244 1.235 1.216 7 2.682 2.148 1.834 1.645 1.498 1.397 1.320 1.276 1.226 1.203 1.177 1.163 6 2.538 2.037 1.749 1.556 1.426 1.331 1.253 1.200 1.163 1.143 1.114 1.113 5 2.349 1.889 1.601 1.426 1.311 1.224 1.152 1.120 1.088 1.053 1.036 1.033 4 2.153 1.738 1.468 1.319 1.206 1.123 1.065 1.017 0.998 0.973 0.965 0.943 3 1.926 1.537 1.309 1.163 1.083 1.004 0.954 0.913 0.878 0.856 0.839 0.832 2 1.726 1.386 1.172 1.049 0.957 0.895 0.847 0.809 0.788 0.765 0.750 0.739 1 1.757 1.468 1.298 1.191 1.105 1.050 1.003 0.982 0.958 0.942 0.938 0.924 Plate 1.839 1.488 1.276 1.140 1.047 0.981 0.931 0.896 0.871 0.849 0.836 0.825 Axial 1.319 1.314 1.315 1.310 1.310 1.308 1.314 1.311 1.319 1.314 1.319 1.320 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.520 0.511 0.506 0.498 0.491 0.488 0.479 0.485 0.489 0.505 0.532 0.595 23 0.452 0.445 0.442 0.441 0.438 0.438 0.437 0.446 0.463 0.495 0.541 0.622 22 0.520 0.520 0.523 0.517 0.514 0.522 0.528 0.537 0.561 0.590 0.641 0.744 21 0.626 0.622 0.614 0.619 0.622 0.621 0.633 0.654 0.683 0.719 0.787 0.910 20 0.731 0.731 0.731 0.726 0.734 0.738 0.754 0.784 0.822 0.878 0.967 1.142 19 0.831 0.822 0.826 0.818 0.834 0.851 0.868 0.902 0.958 1.031 1.147 1.354 18 0.919 0.918 0.926 0.920 0.932 0.942 0.968 1.013 .1.067 1.149 1.289 1.525 17 0.998 0.997 0.995 1.007 1.020 1.044 1.069 1.110 1.169 1.264 1.420 1.684 16 1.077 1.072 1.076 1.082 1.095 1.117 1.144 1.192 1.258 1.366 1.541 1.826 15 1.141 1.138 1.142 1.148 1.156 1.180 1.213 1272 1.349 1.470 1.660 1.977 14 1.190 1.195 1.197 1.196 1.219 1.249 1.284 1.348 1.432 1.563 1.804 2.170 13 1.218 1.222 1.219 1.244 1.263 1.296 1.351 1.407 1.519 1.670 1.914 2.324 12 1.248 1.257 1.255 1.272 1.287 1.317 1.365 1.436 1.541 1.713 1.959 2.396 11 1.264 1264 1.268 1.275 1.292 1.325 1.373 1.453 1.565 1.730 1.987 2.417 10 1.258 1.263 t.261 1.278 1.301 1.320 1.377 1.453 1.561 1.731 1.989 2.441 9 1.245 1.236 1.235 1.238 1.270 1.302 1.352 1.410 1.529 1.704 1.956 2.388 8 1.208 1.199 1.209 1.218 1.236 1.269 1.316 1.383 1.492 1.642 1.895 2.309 7 1.153 1.150 1.163 1.171 1.188 1.225 1.257 1.329 1.436 1.588 1.836 2.223 6 1.108 1.106 1.098 1.110 1.127 1.153 1.199 1.252 1.350 1.498 1.726 2.105 5 1.020 1.015 1.014 1.030 1.046 1.066 1.114 1.179 1.262 1.395 1.605 1.963 4 0.929 0.928 0.924 0.943 0.953 0.973 1.015 1.074 1.155 1.277 1.469 1.808 3 0.835 0.825 0.827 0.837 0.849 0.872 0.906 0.954 1.036 1.160 1.344 1.652 2 0.737 0.739 0.734 0.752 0.765 0.782 0.814 0.860 0.931 1.042 1.205 1.495 1 0.926 0.918 0.915 0.913 0.922 0.929 0.949 0.992 1.051 1.129 1.260 1.506 Plate 0.818 0.816 0.816 0.822 0.832 0.849 0.875 0.916 0.978 1.071 1.218 1.469 Axial 1.310 1.314 1.317 1.319 1.325 1.324 1.334 1.345 1.357 1.370 1.385 1.409 A-6

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-6. MURR fuel element 6 peaking factor for the maximum burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.868 0.758 0.691 0.641 0.588 0.557 0.531 0.512 0.498 0.487 0.475 0.462 23 0.889 0.754 0.662 0.597 0.551 0.509 0.476 0.454 0.435 0.427 0.417 0.406 22 0.992 0.853 0.751 0.682 0.637 0.597 0.561 0.531 0.515 0.502 0.482 0.477 21 1.124 0.976 0.864 0.787 0.726 0.695 0.653 . 0.626 0.613 0.598 0.581 0.575 20 1.216 1.064 0.966 0.892 0.826 0.774 0.740 0.714 0.698 0.676 0.667 0.655 19 1.348 1.194 1.088 0.990 0.931 0.878 0.844 0.812 0.792 0.785 0.769 0.760 18 1.435 1.294 1.182 1.087 1.017 0.968 0.930 0.901 0.873 0.855 0.848 0.838 17 1.588 1.419 1.300 1.198 1.125 1.070 1.035 1.000 0.969 0.948 0.934 0.920 16 1.673 1.509 1.372 1.267 1.191 1.134 1.078 1.050 1.014 0.997 0.989 0.981 15 1.778 1.620 1.463 1.351 1.263 1.205 1.154 1.113 1.086 1.061 1.041 1.033 14 1.834 1.652 1.518 1.400 1.328 1.257 1.199 1.151 1.120 1.095 1.085 1.077 13 1.886 1.720 1.577 1.471 1.375 1.301 1.240 1.205 1.168 1.143 1.125 1.109 12 1.909 1.7213 1.595 1.471 1.389 1.327 1.265 1.213 1.193 1.163 1.151 1.129 11 1.899 1.749 1.598 1.488 1.390 1.323 1.269 1.234 1.205 1.178 . 1.157 1.149 10 1.945 1.771 1.613 1.504 1.405 1.332 1.268 1234 1.199 1.177 1.162 1.146 9 1.945 1.766 1.618 1.484 1.386 1.313 1.262 1.227 1.177 1.159 1.134 1.127 8 1.962 1.765 1.612 1.486 1.379 1.300 1.232 1.197 1.168 1.148 1.131 1.103 7 1.903 1.705 1.544 1.426 1.317 1.249 1.193 1.141 1.116 1.084 1.082 1.060 6 1.888 1.666 1.508 1.360 1.279 1.192 1.132 1.104 1.066 1.048 1.031 1.021 5 1.748 1.534 1.379 1.259 1.179 1.109 1.057 1.012 0.992 0.964 0.946 0.939 4 1.662 1.437 1.287 1.173 1.085 1.030 0.975 0.944 0.917 0.901 0.890 0.878 3 1.462 1.294 1.157 1.053 0.977 0.913 0.872 0.841 0.818 0.798 0.779 0.776 2 1.346 1.173 1.047 0.951 0.882 0.830 0.785 0.758 0.729 0.711 0.703 0.690 1 1.369 1.221 1.129 1.056 1.008 0.968 0.929 0.909 0.891 0.869 0.866 0.857 Plate 1.516 1.353 1.229 1.130 1.056 1.000 0.953 0.921 0.896 0.877 0.863 0.852 Axial 1.250 1.264 1.273 1.286 1.286 1.288 1.286 1.294 1.299 1.298 1.300 1.303 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.449 0.438 0.428 0.422 0.400 0.386 0.368 0.354 0.338 0.326 0.311 0.301 23 0.397 0.385 0.384 0.372 0.358 0.351 0.345 0.336 0.326 0.321 0.312 0.309 22 0.459 0.454 0.449 0.439 0.434 0.421 0.413 0.407 0.393 0.378 0.368 0.362 21 0.561 0.555 0.547 0.539 0.534 0.536 0.532 0.534 0.529 0.540 0.548 0.567 20 0.644 0.644 0.643 0.649 0.641 0.655 0.667 0.679 0.715 0.755 0.807 0.900 19 0.759 0.746 0.757 0.755 0.766 0.783 0.805 0.838 0.882 0.954 1.036 1.171 18 0.830 0.827 0.834 0.836 0.845 0.865 0.898 0.933 0.995 1.058 1.161 1.300 17 0.912 0.905 0.915 0.921 0.938 0.962 0.992 1.042 1.109 1.197 1.317 1.468 16 0.974 0.980 0.983 0.995 1.009 1.040 1.075 1.131 1.193 1.294 1.420 1.580 15 1.033 1.041 1.038 1.058 1.070 1.107 1.142 1.200 1.274 1.389 1.528 1.716 14 1.071 1.071 1.079 1.091 1.120 1.148 1.192 1.253 1.327 1.442 1.582 1.770 13 1.104 1.106 1.115 1.129 1.162 1.190 1.230 1.296 1.381 1.497 1.641 1.823 12 1.130 1.135 1.139 1.151 1.175 1.211 1.261 1.327 1.409 1.528 1.667 1.864 11 1.137 1.147 1.157 1.172 1.195 1.233 1.272 1.344 1.420 1.539 1.683 1.868 10 1.134 1.138 1.153 1.164 1.194 1.213 1.263 1.331 1.424 1.543 1.691 1.884 9 1.124 1.130 1.139 1.152 1.166 1.204 1.252 1.314 1.396 1.515 1.663 1.846 8 1.096 1.109 1.112 1.119 1.143 1.176 1.223 1.280 1.375 1.501 1.642 1.841 7 1.061 1.057 1.056 1.078 1.098 1.127 1.163 1.225 1.307 1.413 1.565 1.756 6 1.013 1.014 1.013 1.035 1.046 1.077 1.116 1.190 1.275 1.376 1.527 1.727 5 0.930 0.930 0.932 0.944 0.971 0.995 1.041 1.094 1.169 1.272 1.409 1.599 4 0.868 0.869 0.875 0.881 0.898 0.929 0.962 1.011 1.087 1.184 1.324 1.520 3 0.763 0.774 0.773 0.781 0.802 0.820 0.860 0.909 0.974 1.061 1.187 1.362 2 0.689 0.690 0.696 0.701 0.715 0.745 0.772 0.812 0.874 0.964 1.085 1.263 1 0.854 0.847 0.854 0.854 0.864 0.881 0.896 0.934 0.979 1.038 1.125 1.270 Plate 0.845 0.845 0.848 0.855 0.867 0.888 0.915 0.957 1.013 1.090 1.192 1.331 Axial 1.300 1.311 1.318 1.325 1.331 1.342 1.342 1.357 1.359 1.368 1.371 1.367 A-7

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-7. MURR fuel element 7 peaking factor for the maximum burnup c o r e - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.906 0.766 0.682 0.619 0.573 0.539 0.521 0.497 0.479 0.465 0.448 0.434 23 0.925 0.756 0.655 0.586 0.533 0.488 0.458 0.437 0.415 0.402 0.382 0.377 22 1.045 0.867 0.758 0.670 0.608 0.569 0.529 0.507 0.482 0.463 0.456 0.442 21 1.199 1.005 0.873, 0.783 0.714 0.666 0.636 0.594 0.574 0.558 0.540 0.536 20 1.342 1.127 0.983 0.877 0.812 0.755 0.714 0.683 0.659 0.639 0.628 0.604 19 1.501 1.260 1.113 1.001 0.912 0.859 0.818 0.783 0.754 0.734 0.713 0.698 18 1.627 1.385 1.214 1.097 1.016 0.950 0.889 0.861 0.834 0.802 0.783 0.773 17 1.802 1.526 1.349 1.205 1.114 1.049 0.999 0.960 0.916 0.891 0.875 0.859 16 1.941 1.648 1.458 1.304 1.199 1.136 1.077 1.028 0.996 0.971 0.947 0.935 15 2.072 1.759 1.539 1.400 1.294 1.204 1.144 1.104 1.067 1.035 1.016 1.001 14 2.139 1.831 1.619 1.459 1.356 1.274 1.208 1.163 1.121 1.091 1.067 1.056 13 2.215 1.906 1.686 1.517 1.405 1.323 1.256 1.204 1.157 1.139 1.113 1.095 12 2.258 1.947 1.721 1.560 1.428 1.351 1.292 1.238 1.188 1.155 1.136 1.127 11 2.266 1.976 1.738 1.589 1.459 1.376 1.302 1.252 1.204 1.178 1.147 1.144 10 2.303 1.972 1.744 1.581 1.462 1.382 1.313 1.252 1.215 1.187 1.167 1.139 9 2.296 1.952 1.736 1.566 1.451 1.350 1.292 1.237 1.203 1.182 1.143 1.133 8 2.283 1.942 1.717 1.541 1.436 1.347 1.271 1.221 1.179 1.151 1.127 1.108 7 2.208 1.881 1.663 1.492 1.381 1.301 1.230 1.168 1.136 1.102 1.085 1.062 6 2.137 1.793 1.574 1.426 1.318 1.229 1.169 1.123 1.081 1.055 1.026 1.011 5 2.001 1.677 1.470 1.319 1.216 1.135 1.076 1.035 1.008 0.971 0.959 0.938 4 1.848 1.549 1.343 1.215 1.116 1.042 0.995 0.941 0.918 0.899 0.875 0.870 3 1.655 1.383 1.206 1.083 0.997 0.931 0.884 0.845 0.817 0.798 0.780 0.776 2 1.511 1.255 1.081 0.974 0.898 0.831 0.789 0.768 0.745 0.716 0.700 0.693 1 1.538 1.311 1.184 1.092 1.028 0.993 0.936 0.912 0.895 0.892 0.874 0.857 Plate 1.759 1.492 1.313 1.184 1.093 1.026 0.973 0.933 0.902 0.878 0.859 0.845 Axial 1.285 1.300 1.304 1.317 1.313 1.322 1.324 1.318 1.323 1.327 1.334 1.329 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.422 0.412 0.402 0.389 0.382 0.365 0.357 0.349 0.335 0.336 0.330 0.337 23 0.365 0.356 0.356 0.346 0.335 0.330 0.326 0.324 0.322 0.320 0.326 0.340 22 0.430 0.422 0.418 0.407 0.405 0.393 0.390 0.382 0.384 0.384 0.388 0.407 21 0.521 0.508 0.497 0.499 0.481 0.481 0.473 0.476 0.482 0.490 0.512 0.544 20 0.598 0.591 ' 0.583 0.573 0.569 0.577 0.583 0.593 0.613 0.637 0.678 0.764 19 0.687 0.676 0.675 0.675 0.672 0.679 0.683 0.703 0.723 0.773 0.850 0.960 18 0.760 0.747 0.749 0.751 0.745 0.754 0.760 0.788 0.821 0.867 0.945 1.075 17 0.841 0.831 0.823 0.830 0.839 0.842 0.854 0.880 0.927 0.986 1.086 1.227 16 0.916 0.909 0.903 0.895 0.915 0.927 0.943 0.981 1.031 1.109 1.220 1.397 15 0.989 0.986 0.977 0.981 1.000 1.011 1.054 1.098 1.178 1.268 1.415 1.649 14 1.046 1.044 1.028 1.049 1.058 1.082 1.117 1.163 1.254 1.367 1.531 1.776 13 1.083 1.081 1.073 1.083 1.105 1.140 1.176 1.237 1.327 1.452 1.620 1.882 12 1.127 1.119 1.109 1.122 1.146 1.175 1.216 1.283 1.355 1.487 1.666 1.939 11 1.135 1.132 1.128 1.138 1.163 1.188 1.230 1.295 1.384 1.517 1.692 1.957 10 1.133 1.133 1.136 1.151 1.158 1.189 1.233 ' 1.298 1.381 1.504 1.696 1.981 9 1.117 1.105 1.116 1.128 1.149 1.173 1.219 1.277 1.349 1.490 1.671 1.935 8 1.094 1.086 1.092 1.101 1.110 1.146 1.188 1247 1.334 1.455 1.623 1.892 7 1.056 1.036 1.052 1.058 1.075 1.103 1.141 1.198 1.283 1.391 1.561 1.821 6 1.006 0.999 1.001 1.012 1.027 1.044 1.083 1.147 1.216 1.329 1.496 1.757 5 0.935 0.931 0.934 0.945 0.955 0.972 1.005 1.059 1.133 1.239 1.387 1.642 4 0.856 0.857 0.855 0.861 0.877 0.904 0.939 0.982 1.053 1.147 1.301 1.540 3 0.770 0.765 0.772 0.771 0.786 0.799 0.835 0.878 0.943 1.046 1.186 1.404 2 0.685 0.684 0.686 0.689 0.702 0.720 0.758 0.800 0.859 0.946 1.083 1.305 1 0.852 0.845 0.837 0.847 0.859 0.867 0.884 0.922 0.964 1.035 1.142 1.325 Plate 0.835 0.828 0.826 0.830 0.839 0.853 0.877 0.914 0.967 1.046 1.162 1.344 Axial 1.333 1.342 1.349 1.360 1.360 1.368 1.379 1.393 1.404 1.424 1.433 1.447 A-8

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design fqr Once-Through Operation 30441 R00031 /B Table A-8. MURR fuel element 8 peaking factor for the maximum burnup core - )

Axial Fuel plate number node 2 3 4 5 6 7 8 9 10 11 12 24 0.760 0.663 0.605 0.553 0.518 0.486 0.462 0.445 0.429 0.415 0.400 0.390 23 0.783 0.665 0.586 0.515 0.475 0.435 0.406 0.386 0.372 0.363 0.345 0.334 22 0.867 0.749 0.657 0.596 0.538 0.504 0.471 0.447 0.423 0.409 0.396 0.383 21 1.003 0.865 0.771 0.696 0.636 0.594 0.558 0.530 0.502 0.482 0.464 0.454 20 1.070 0.939 0.847 0.764 0.713 0.657 0.614 0.589 0.571 0.539 0.522 0.508 19 1.197 1.057 0.936 0.858 0.796 0.737 0.709 0.672 0.648 0.621 0.598 0.584 18 1.261 1.143 1.033 0.936 0.881 0.827 0.781 0.749 0.709 0.675 0.663 0.645 17 1.412 1.259 1.152 1.057 0.981 0.913 0.865 0.828 0.799 0.770 0.744 0.718 16 1.463 1.340 1.220 1.128 1.057 0.996 0.939 0.902 0.860 0.836 0.811 0.791 15 1.586 1.438 1.321 1.212 1.142 1.064 1.018 0.975 0.944 0.910 0.885 0.873 14 1.621 1.491 1.384 1.280 1.203 1.132 1.072 1.038 0.998 0.964 0.947 0.932 13 1.691 1.553 1.425 1.326 1.245 1.178 1.125 1.082 1.046 1.015 0.995 0.990 12 1.713 1.574 1.455 1.354 1.275 1.203 1.149 1.114 1.085 1.058 1.038 1.027 11 1.732 1.609 1.484 1.373 1.298 1.226 1.173 1.140 1.098 1.085 1.057 1.048 10 1.753 1.636 1.502 1.387 1.300 1.241 1.195 1.150 1.113 1.093 1.069 1.062 9 1.761 1.627 1.499 1.393 1.303 1.242 1.176 1.136 1.106 1.079 1.058 1.046 8 1.787 1.620 1.486 1.377 1.307 1.223 1.169 1.123 1.080 1.061 1.053 1.034 7 1.734 1.584 1.443 1.332 1.250 1.181 1.123 1.085 1.059 1.029 1.009 0.990 6 1.731 1.553 1.415 1.303 1.218 1.145 1.078 1.039 1.012 0.997 0.969 0.962 5 1.615 1.440 1.307 1.200 1.124 1.063 1.008 0.958 0.941 0.911 0.892 0.878 4 1.557 1.381 1.231 1.141 1.059 0.984 0.935 0.904 0.883 0.854 0.847 0.828 3 1.380 1.218 1.102 0.998 0.931 0.864 0.829 0.797 0.770 0.750 0.744 0.736 2 1.268 1.113 0.996 0.910 0.851 0.793 0.742 0.710 0.695 0.676 0.667 0.662 1 1.303 1.176 1.085 1.007 0.963 0.917 0.883 0.874 0.854 0.835 0.818 0.819 Plate 1.623 1.463 1.332 1.225 1.147 1.077 1.024 0.985 0.953 0.926 0.905 0.891 Axial 1.260 1.279 1.290 1.301 1.304 1.319 1.335 1.335 1.336 1.350 1.351 1.364 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.374 0.365 0.344 0.336 0.318 0.308 0.293 0.280 0.263 0.249 0.233 0.217 23 0.323 0.308 0.305 0.295 0.281 0.273 0.264 0.256 0.244 0.234 0.222 0.213 22 0.371 0.359 0.349 0.339 0.324 0.315 0.305 0.287 0.277 0.265 0.248 0.229 21 0.438 0.420 0.407 0.394 0.384 0.372 0.357 0.343 0.328 0.307 0.288 0.269 20 0.493 0.478 0.455 0.444 0.434 0.416. 0.401 0.384 0.364 0.341 0.313 0.285 19 0.568 0.540 0.529 0.511 0.496 0.481 0.466 0.444 0.423 0.393 0.364 0.334 18 0.618 0.595 0.578 0.563 0.548 0.525 0.508 0.483 0.460 0.427 0.394 0.354 17 0.701 0.678 0.657 0.638 0.616 0.601 0.575 0.557 0.538 0.510 0.470 0.430 16 0.770 0.757 0.737 0.722 0.716 0.700 0.681 0.675 0.657 0.642 0.625 0.603 15 0.857 0.841 0.833 0.828 0.832 0.835 0.845 0.864 0.888 0.930 0.976 1.022 14 0.919 0.912 0.913 0.917 0.933 0.946 0.966 1.000 1.046 1.118 1.204 1.297 13 0.974 0.975 0.973 0.987 1.005 1.016 1.061 1.098 1.149 1.229 1.324 1.440 12 1.013 1.021 1.012 1.021 1.034 1.062 1.095 1.142 1.207 1.281 1.374 1.502 11 1.030 1.025 1.034 1.050 1.061 1.082 1.114 1.175 1.234 1.311 1.419 1.546 10 1.049 1.040 1.047 1.059 1.075 1.089 1.141 1.180 1.255 1.347 1.455 1.576 9 1.037 1.031 1.035 1.046 1.063 1.090 1.126 1.173 1.242 1.333 1.432 1.564 8 1.030 1.023 1.029 1.034 1.050 1.081 1.118 1.176 1.239 1.333 1.445 1.590 7 0.982 0.986 0.987 0.996 1.011 1.047 1.080 1.128 1.201 1.290 1.402 1.538 6 0.946 0.951 0.949 0.959 0.975 1.007 1.037 1.088 1.154 1.246 1.365 1.532 5 0.873 0.874 0.879 0.889 0.907 0.921 0.958 1.011 1.074 1.167 1.278 1.426 4 0.824 0.811 0.818 0.831 0.857 0.868 0.906 0.947 1.009 1.100 1.222 1.379 3 0.732 0.732 0.740 0.746 0.754 0.777 0.809 0.846 0.914 1.000 1.104 1.253 2 0.661 0.664 0.659 0.671 0.680 0.704 0.733 0.774 0.824 0.905 1.024 1.170 1 0.819 0.802 0.799 0.814 0.822 0.832 0.854 0.882 0.920 0.981 1.065 1.188 Plate 0.877 0.867 0.861 0.862 0.866 0.874 0.891 0.915 0.949 0.998 1.060 1.142 Axial 1.368 1.373 1.391 1.406 1.419 1.425 1.464 1.475 1.513 1.544 1.570 1.593 A-9

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-9. MURR fuel element 1 peaking factor for the extreme burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.846 0.699 0.602 0.545 0.513 0.472 0.449 0.426 0.409 0.394 0.383 0.370 23 0.864 0.688 0.589 0.511 0.466 0.434 0.393 0.369 0.351 0.338 0.323 0.313 22 0.994 0.799 0.666 0.588 0.525 0.484 0.457 0.429 0.410 0.391 0.371 0.361 21 1.137 0.925 0.775 0.680 0.614 0.560 0.530 0.502 0.476 0.459 0.436 0.427 20 1.295 1.024 0.877 0.772 0.702 0.649 0.604 0.576 0.547 0.529 0.518 0.498 19 1.467 1.168 0.999 0.871 0.789 0.739 0.699 0.656 0.629 0.602 0.575 0.550 18 1.647 1.299 1.112 0.987 0.895 0.831 0.771 0.728 0.700 0.668 0.637 0.614 17 1.802 1.447 1.233 1.090 0.988 0.902 0.853 0.802 0.773 0.738 0.701 0.689 16 1.991 1.592 1.351 1.179 1.068 0.995 0.929 0.873 0.824 0.803 0.782 0.754 15 2.154 1.727 1.464 1.296 1.164 1.072 1.003 0.937 0.899 0.870 0.837 0.807 14 2.317 1.862 1.567 1.375 1.244 1.152 1.073 1.003 0.970 0.928 0.894 0.875 13 2.446 1.960 1.681 1.475 1.336 1.229 1.149 1.095 1.033 1.003 0.961 0.928 12 2.553 2.067 1.761 1.544 1.396 1.294 1.213 1.155 1.104 1.068 1.031 1.001 11 2.641 2.152 1.840 1.628 1.475 1.371 1.274 1.223 1.175 1.122 1.085 1.063 10 2.794 2.226 1.888 1.696 1.544 1.427 1.349 1.278 1.224 1.185 1.149 1.131 9 2.886 2.317 1.990 1.754 1.601 1.496 1.394 1.334 1.282 1.246 1.205 1.180 8 2.945 2.364 2.006 1.785 1.620 1.505 1.423 1.355 1.315 1.276 1.241 1.217 7 2.930 2.353 1.996 1.777 1.619 1.512 1.431 1.356 1.310 1.266 1.231 1.217 6 2.852 2.293 1.962 1.739 1.587 1.478 1.385 1.334 1.285 1.253 1.219 1.193 5 2.722 2.209 1.877 1.646 1.521 1.394 1.330 1.265 1.229 1.191 1.164 1.152 4 2.541 2.041 1.732 1.538 1.400 1.306 1.241 1.187 1.137 1.097 1.080 1.070 3 2.282 1.850 1.581 1.400 1.256 1.171 1.117 1.067 1.035 1.009 0.974 0.965 2 2.072 1.678 1.422 1.255 1.155 1.083 1.019 0.968 0.937 0.920 0.892 0.886 1 2.175 1.814 1.581 1.443 1.355 1.287 1.226 1.198 1.158 1.145 1.115 1.098 Plate 2.184 1.759 1.499 1.326 1.207 1.121 1.055 1.003 0.963 0.933 0.902 0.883 Axial 1.404 1.399 1.393 1.401 1.397 1.405 1.413 1.408 1.420 1.425 1.432 1.435 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.359 0.346 0.334 0.323 0.311 0.309 0.291 0.282 0.269 0.260 0.257 0.259 23 0.308 0.298 0.286 0.280 0.270 0.262 0.260 0.250 0.245 0.239 0.238 0.242 22 0.353 0.348 0.338 0.325 0.317 0.307 0.299 0.295 0.286 0.280 0.282 0.283 21 0.417 0.404 0.387 0.382 0.376 0.360 0.350 0.341 0.333 0.325 0.322 0.326 20 0.481 0.457 0.450 0.436 0.423 0.410 0.404 0.392 0.387 0.377 0.370 0.376 19 0.535 0.517 0.502 0.485 0.477 0.470 0.452 0.440 0.429 0.421 0.418 0.416 18 0.593 0.577 0.573 0.563 0.536 0.524 0.505 0.499 0.478 0.468 0.464 0.463 17 0.668 0.642 0.632 0.614 0.597 0.580 0.561 0.546 0.529 0.518 0.513 0.514 16 0.735 0.700 0.682 0.676 0.646 0.634 0.615 0.595 0.582 0.570 0.558 0.568 15 0.795 0.770 0.739 0.716 0.702 0.687 0.668 0.643 0.633 0.614 0.605 0.613 14 0.851 0.829 0.804 0.774 0.756 0.732 0.716 0.700 0.686 0.669 0.656 0.661 13 0.908 0.889 0.855 0.837 0.821 0.798 0.771 0.752 0.738 0.722 0.713 0.724 12 0.967 0.945 0.923 0.903 0.876 0.853 0.834 0.816 0.802 0.791 0.785 0.804 11 1.027 1.001 0.975 0.956 0.941 0.932 0.914 0.901 0.892 0.893 0.907 0.958 10 1.099 1.081 1.063 1.051 1.040 1.030 1.036 1.052 1.080 1.113 1.197 1.346 9 1.159 1.146 1.135 1.132 1.142 1.145 1.163 1.209 1.274 1.358 1.524 1.802 8 1.201 1.182 1.178 1.179 1.193 1.216 1.247 1.297 1.370 1.487 1.678 2.013 7 1.201 1.204 1.196 1.196 1.206 1.231 1.258 1.307 1.399 1.530 1.728 2.081 6 1.178 1.168 1.174 1.178 1.181 1.203 1.242 1.290 1.389 1.513 1.725 2.083 5 1.127 1.118 1.123 1.130 1.134 1.166 1.184 1.240 1.333 1.470 1.671 2.017 4 1.053 1.042 1.045 1.057 1.061 1.082 1.115 1.170 1.244 1.375 1.580 1.920 3 0.954 0.952 0.951 0.956 0.966 0.980 1.018 1.065 1.147 1.264 1.453 1.771 2 0.869 0.865 0.859 0.868 0.881 0.899 0.927 0.975 1.046 1.161 1.340 1.638 1 1.090 1.082 1.073 1.077 1.071 1.091 1.113 1.139 1.211 1.291 1.427 1.690 Plate 0.864 0.848 0.836 0.828 0.821 0.820 0.822 0.833 0.858 0.898 0.972 1.109 Axial 1.447 1.478 1.489 1.504 1.530 1.563 1.594 1.635 1.697 1.773 1.850 1.955 A-10

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-10. MURR fuel element 2 peaking factor for the extreme burnup core ( -

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.713 0.610 0.546 0.509 0.470 0.444 0.419 0.406 0.394 0.383 0.363 0.351 23 0.724 0.619 0.535 0.471 0.429 0.404 0.367 0.351 0.335 0.324 0.306 0.298 22 0.811 0.689 0.604 0.545 0.494 0.463 0.436 0.404 0.386 0.374 0.360 0.353 21 0.930 0.798 0.696 0.633 0.573 0.533 0.502 0.480 0.457 0.440 0.418 0.406 20 1.007 0.881 0.774 0.705 0.649 0.597 0.569 0.538 0.516 0.489 0.473 0.461 19 1.129 0.988 0.881 0.795 0.731 0.680 0.641 0.607 0.584 0.552 0.538 0.526 18 1.226 1.084 0.966 0.883 0.800 0.754 0.704 0.675 0.644 0.626 0.601 0.585 17 1.379 1.215 1.087 0.980 0.904 0.843 0.791 0.755 0.725 0.693 0.673 0.646 16 1~458 1.297 1.161 1.065 0.974 0.914 0.857 0.816 0.782 0.746 0.719 0.701 15 1.591 1.404 1.263 1.150 1.060 0.988 0.941 0.891 0.847 0.820 0.786 0.759 14 1.688 1.491 1.341 1.223 1.131 1.061 1.002 0.946 0.911 0.873 0.839 0.809 13 1.802 1.589 1.430 1.313 1.207 1.135 1.071 1.020 0.980 0.943 0.903 0.878 12 1.871 1.662 1.499 1.378 1.284 1.203 1.134 1.074 1.039 0.992 0.956 0.924 11 1.950 1.740 1.579 1.440 1.334 1.259 1.188 1.136 1.087 1.043 1.022 0.984 10 2.045 1.808 1.644 1.501 1.396 1.310 1.252 1.177 1.133 1.116 1.077 1.053 9 2.097 1.869 1.686 1.544 1.448 1.347 1.291 1.226 1.186 1.155 1.131 1.108 8 2.188 1.933 1.748 1.595 1.485 1.390 1.324 1.272 1.229 1.189 1.167 1.139 7 2.149 1.899 1.727 1.581 1.465 1.375 1.318 1.251 1.224 1.189 1.166 1.144 6 2.173 1.926 1.717 1.557 1.448 1.370 1.299 1.250 1.202 1.175 1.150 1.139 5 2.068 1.802 1.624 1.478 1.370 1.287 1.231 1.189 1.146 1.117 1.093 1.073 4 2.001 1.735 1.546 1.415 1.307 1.235 1.180 1.121 1.073 1.056 1.034 1.016 3 1.809 1.566 1.388 1.262 1.172 1.115 1.058 1.012 0.976 0.951 0.927 0.919 2 1.686 1.451 1.302 1.170 1.083 1.010 0.966 0.930 0.886 0.864 0.850 0.831 1 1.731 1.539 1.410 1.318 1.245 1.186 1.143 1.111 1.092 1.065 1.053 1.035 Plate 1.840 1.617 1.452 1.324 1.226 1.151 1.092 1.042 1.003 0.971 0.944 0.921 Axial 1.374 1.381 1.392 1.392 1.399 1.395 1.401 1.411 1.416 1.415 1.428 1.435 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.344 0.327 0.320 0.300 0.296 0.284 0.271 0.252 0.241 0.227 0.215 0.199 23 0.292 0.283 0.277 0.267 0.258 0.250 0.241 0.229 0.217 0.209 0.198 0.191 22 0.339 0.328 0.316 0.307 0.303 0.287 0.280 0.265 0.248 0.237 0.226 0.206 21 0.401 0.391 0.375 0.365 0.356 0.347 0.326 0.312 0.293 0.280 0.266 0.245 20 0.449 0.440 0.423 0.409 0.397 0.381 0.363 0.347 0.332 0.310 0.286 0.264 19 0.514 0.493 0.477 0.466 0.452 0.429 0.411 0.389 0.373 0.353 0.326 0.297 18 0.561 0.541 0.524 0.515 0.486 0.472 0.452 0.433 0.409 0.380 0.350 0.319 17 0.630 0.602 0.583 0.562 0.547 0.528 0.507 0.483 0.451 0.421 0.387 0.354 16 0.681 0.656 0.629 0.617 0.588 0.564 0.544 0.521 0.493 0.456 0.418 0.376 15 0.736 0.712 0.693 0.672 0.644 0.616 0.595 0.568 0.536 0.496 0.461 0.412 14 0.788 0.761 0.737 0.720 0.689 0.673 0.639 0.604 0.570 0.535 0.490 0.435 13 0.848 0.820 0.796 0.771 0.744 0.718 0.689 0.650 0.611 0.570 0.529 0.477 12 0.897 0.872 0.856 0.832 0.801 0.772 0.736 0.705 0.664 0.623 0.569 0.524 11 0.952 0.938 0.913 0.888 0.855 0.842 0.819 0.779 0.756 0.712 0.675 0.627 10 1.028 1.014 0.996 0.979 0.963 0.959 0.952 0.948 0.948 0.950 0.965 0.987 g 1.084 1.071 1.069 1.074 1.084 1.083 1.101 1.120 1.173 1.233 1.310 1.425 8 1.123 1.115 1.126 1.128 1.138 1.163 1.193 1.237 1.298 1.380 1.498 1.656 7 1.128 1.134 1.126 1.132 1.143 1.170 1.205 1.246 1.319 1.424 1.555 1.736 6 1.120 1.115 1.109 1.119 1.139 1.159 1.207 1.255 1.336 1.445 1.579 1.787 5 1.072 1.058 1.061 1.065 1.082 1.110 1.146 1.198 1.277 1.373 1~517 1.720 4 1.006 1.001 1.000 1.003 1.026 1.052 1.099 1.143 1.218 1.318 1.471 1.661 3 0.911 0.905 0.907 0.908 0.922 0.947 0.990 1.043 1.111 1.216 1.355 1.557 2 0.823 0.825 0.829 0.839 0.853 0.873 0.910 0.958 1.028 1.119 1.249 1.449 1 1.025 1.017 1.023 1.029 1.043 1.051 1.077 1.104 1.152 1.230 1.332 1.505 Plate 0.903 0.887 0.875 0.865 0.857 0.854 0.854 0.856 0.869 0.890 0.926 0.982 Axial 1.444 1.478 1.488 1.513 1.540 1.583 1.632 1.694 1.776 1.875 1.971 2.102 A-11

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00031/B Table A-11. MURR fuel element 3 peaking factor for the extreme burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.794 0.672 0.591 0.539 0.505 0.474 0.451 0.433 0.423 0.409 0.392 0.379 23 0.817 0.678 0.577 0.515 0.468 0.437 0.399 0.372 0.367 0.354 0.337 0.325 22 0.929 0.774 0.666 0.589 0.533 0.492 0.466 0.446 0.428 0.412 0.393 0.382 21 1.059 0.876 0.764 0.677 0.620 *o.579 0.541 0.512 0.493 0.475 0.462. 0.450 20 1.158 0.977 0.850 0.760 0.707 0.657 0.615 0.588 0.566 0.549 0.533 0.511 19 1.283 1.089 0.955 0.861 0.800 0.749 0.696 0.666 0.645 0.618 0.600 0.581 18 1.410 1.204 1.059 0.953 0.889 0.821 0.779 0.741 0.706 0.691 0.671 0.656 17 1.615 1.349 1.192 1.086 1.005 0.937 0.870 0.837 0.812 0.781 0.753 0.733 16 1.763 1.507 1.329 1.202 1.106 1.030 0.971 0.931 0.894 0.858 0.834 0.809 15 1.956 1.649 1.452 1.318 1.211 1.133 1.070 1.024 0.984 0.945 0.919 0.897 14 2.110 1.782 1.571 1.417 1.308 1.217 1.145 1.096 1.061 1.008 1.006 0.979 13 2.247 1.903 1.670 1.503 1.391 1.291 1.226 1.177 1.134 1.103 1.086 1.057 12 2.338 1.999 1.766 1.591 1.482 1.379 1.300 1.239 1.196 1.161 1.141 1.106 11 2.428 2.060 1.838 1.647 1.526 1.436 1.352 1.303 1.261 1.217 1.186 1.176 10 2.535 2.150 1.898 1.721 1.574 1.473 1.407 1.354 1.294 1.255 1.232 1.215 9 2.562 2.178 1.913 1.731 1.613 1.500 1.435 1.380 1.329 1.295 1.262 1.236 8 2.628 2.225 1.959 1.767 1.622 1.524 1.440 1.397 1.340 1.311 1.272 1.249 7 2.586 2.182 1.913 1.725 1.598 1.507 1.422 1.368 1.314 1.286 1.269 1.252 6 2.526 2.148 1.882 1.697 1.555 1.464 1.389 1.336 1.286 1.251 1.232 1.208 5 2.401 2.032 1.778 1.599 1.484 1.386 1.337 1.277 1.222 1.195 1.170 1.137 4 2.285 1.903 1.667 1.491 1.377 1.297 1.225 1.176 1.145 1.114 1.078 1.070 3 2.070 1.735 1.506 1.354 1.255 1.168 1.092 1.056 1.031 1.000 0.971 0.960 2 1.911 1.596 1.386 1.235 1.144 1.061 1.008 0.966 0.927 0.897 0.889 0.881 1 1.972 1.707 1.511 1.398 1.310 1.243 1.209 1.170 1.137 1.120 1.099 1.086 Plate 1.873 1.584 1.391 1.254 1.159 1.084 1.025 0.984 0.949 0.920 0.899 0.880 Axial 1.390 1.392 1.396 1.396 1.386 1.393 1.391 1.406 1.399 1.411 1.401 1.408 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.369 0.358 0.344 0.329 0.317 0.310 0.297 0.285 0.280 0.273 0.266 0.268 23 0.317 0.310 0.298 0.289 0.281 0.277 0.270 0.268 0.260 0.255 0.260 0.266 22 0.372 0.359 0.351 0.338 0.330 0.324 0.315 0.309 0.306 0.300 0.297 0.304 21 0.427 0.426 0.415 0.403 0.391 0.382 0.378 0.362 0.360 0.355 0.356 0.354 20 0.498 0.484 0.474 0.458 0.448 0.441 0.430 0.422 0.414 0.404 0.404 0.400 19 0.564 0.547 0.532 0.530 0.517 0.506 0.492 0.479 0.476 0.468 0.459 0.461 18 0.637 0.621 0.613 0.599 0.581 0.569 0.555 0.538 0.524 0.517 0.511 0.513 17 0.714 0.693 0.679 0.665 0.656 0.637 0.620 0.614 0.597 0.587 0.576 0.581 16 0.795 0.775 0.754 0.734 0.721 0.716 0.700 0.691 0.678 0.663 0.663 0.670 15 0.877 0.859 0.845 0.823 0.821 0.817 0.814 0.808 0.815 0.828 0.855 0.896 14 0.963 0.946 0.931 0.919 0.920 0.918 0.917 0.924 0.950 0.991 1.052 1.148 13 1.035 1.018 1.012 1.001 1.002 1.006 1.015 1.030 1.056 1.108 1.187 1.317 12 1.091 1.087 1.067 1.067 1.068 1.087 1.081 1.110 1.136 1.213 1.302 1.446 11 1.152 1.131 1.129 1.120 1.124 1.134 1.158 1.185 1.224 1.298 1.402 1.573 10 1.192 1.181 1.170 1.175 1.186 1.194 1.233 1.275 1.333 1.421 1.561 1.768 9 1.227 1.220 1.213 1.224 1.232 1.245 1.282 1.348 1.421 1.527 1.690 1.942 8 1.236 1.232 1.234 1.244 1.260 1.282 1.322 1.380 1.445 1.565 1.747 2.014 7 1.241 1.225 1.222 1.228 1.238 1.267 1.299 1.360 1.440 1.556 1.739 2.000 6 1.204 1.194 1.189 1.196 1.206 1.224 1.267 1.318 1.408 1.533 1.715 1.986 5 1.134 1.123 1.126 1.132 1.145 1.166 1.197 1.259 1.334 1.450 1.628 1.900 4 1.059 1.056 1.053 1.058 1.066 1.095 1.122 1.181 1.258 1.375 1.553 1.813 3 0.947 0.944 0.942 0.947 0.961 0.981 1.013 1.071 1.147 1.252 1.416 1.661 2 0.877 0.872 0.858 0.865 0.882 0.901 0.936 0.971 1.046 1.157 1.304 1.557 1 1.072 1.055 1.058 1.057 1.062 1.077 1.096 1.124 1.186 1.272 1.391 1.589 Plate 0.867 0.855 0.846 0.842 0.843 0.848 0.859 0.880 0.912 0.964 1.045 1.173 Axial 1.418 1.428 1.444 1.463 1.482 1.497 1.525 1.554 1.570 1.607 1.655 1.700 A-12

~--------------------- - - - - - -

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Table A-12. MURR fuel element 4 peaking factor for the extreme burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.713 0.635 0.580 0.546 0.503 0.476 0.455 0.449 0.430 0.413 0.400 0.390 23 0.748 0.639 0.569 0.515 0.467 0.431 0.411 0.393 0.375 0.364 0.348 0.338 22 0.831 0.715 0.638 0.580 0.537 0.498 0.477 0.451 0.431 0.417 0.407 0.393 21 0.938 0.833 0.735 0.683 0.630 *o.588 0.556 0.530 0.511 0.494 0.482 0.463 20 0.995 0.891 0.807 0.740 0.694 0.650 0.619 0.594 0.570 0.556 0.540 0.526 19 1.111 0.992 0.909 0.847 0.793 0.757 0.719 0.683 0.657 0.641 0.621 0.607 18 1.170 1.073 0.996 0.925 0.868 0.823 0.781 0.746 0.725 0.714 0.695 0.672 17 1.340 1.226 1.129 1.045 0.982 0.930 0.891 0.861 0.833 0.804 0.782 0.760 16 1.442 1.325 1.228 1.136 1.072 1.022 0.973 0.938 0.904 0.885 0.869 0.840 15 1.583 1.464 1.359 1.267 1.199 1.127 1.082 1.044 1.010 0.993 0.972 0.946 14 1.661 1.541 1.439 1.350 1.278 1.215 1.166 1.128 1.094 1.069 1.049 1.029 13 1.792 1.666 1.554 1.443 1.381 1.316 1.253 1.210 1.168 1.156 1.135 1.128 12 1.844 1.717 1.617 1.505 1.430 1.366 1.310 1.274 1.242 1.204 1.199 1.179 11 1.925 1.804 1.682 1.581 1.482 1.428 1.369 1.337 1.294 1.267 1.245 1.233 10 2.003 1.873 1.749 1.647 1.541 1.475 1.415 1.367 1.332 1.292 1.274 1.269 9 2.053 1.906 1.771 1.655 1.567 1.482 1.406 1.367 1.339 1.309 1.290 1.269 8 2.095 1.937 1.785 1.669 1.580 1.492 1.436 1.398 1.349 1.323 1.297 1.280 7 2.083 1.912 1.754 1.630 1.538 1.447 1.388 1.356 1.315 1.276 1.274 1.253 6 2.099 1.913 1.746 1.614 1.512 1.441 1.377 1.320 1.282 1.261 1.242 1.233 5 1.999 1.785 1.626 1.516 1.418 1.339 1.272 1.243 1.211 1.178 1.164 1.146 4 1.933 1.713 1.562 1.437 1.338 1.260 1.207 1.169 1.133 1.104 1.091 1.082 3 1.741 1.550 1.415 1.303 1.216 1.139 1.088 1.055 1.019 0.996 0.983 0.978 2 1.643 1.459 1.306 1.196 1.105 1.044 1.000 0.967 0.930 0.907 0.886 0.883 1 1.692 1.526 1.433 1.346 1.273 1.218 1.175 1.143 1.119 1.103 1.089 1.077 Plate 1.530 1.393 1.283 1.192 1.120 1.061 1.015 0.982 0.951 0.929 0.913 0.898 Axial 1.346 1.364 1.365 1.373 1.384 1.379 1.388 1.397 1.391 1.397 1.394 1.398 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.378 0.369 0.356 0.346 0.337 0.317 0.307 0.291 0.276 0.259 0.247 0.232 23 0.327 0.319 0.312 0.305 0.296 0.285 0.279 0.269 0.256 0.247 0.234 0.223 22 0.382 0.373 0.356 0.348 0.343 0.330 0.319 0.309 0.295 0.283 0.265 0.247 21 0.445 0.435 0.428 0.414 0.405 0.396 0.384 0.370 0.351 0.333 0.315 0.297 20 0.509 0.495 0.482 0.471 0.464 0.449 0.431 0.419 0.395 0.375 0.352 0.319 19 0.593 0.577 0.553 0.548 0.535 0.513 0.499 0.485 0.460 0.438 0.410 0.373 18 0.659 0.640 0.626 0.610 0.601 0.572 0.557 0.534 0.505 0.479 0.443 0.402 17 0.736 0.716 0.703 0.692 0.682 0.662 0.644 0.612 0.584 0.550 0.511 0.466 16 0.822 0.807 0.800 0.780 0.764 0.743 0.722 0.708 0.684 0.649 0.615 0.564 15 0.939 0.922 0.904 0.895 0.893 0.886 0.893 0.891 0.886 0.895 0.895 0.907 14 1.024 1.027 1.006 1.013 1.021 1.034 1.044 1.074 1.094 1.137 1.197 1.262 13 1.107 1.108 1.106 1.111 1.123 1.138 1.177 1.213 1.263 1.317 1.388 1.481 12 1.172 .1.167 1.169 1.180 1.203 1.227 1.265 1.297 1.358 1.428 1.508 1.610 11 1.226 1.221 1.218 1.230 1.254 1.280 1.320 1.366 1.432 1.515 1.614 1.741 10 1.250 1.253 1.264 1.279 1.299 1.317 1.374 1.422 1.511 1.613 1.717 1.855 9 1.265 1.263 1.278 1.291 1.316 1.351 1.398 1.464 1.540 1.619 1.744 1.894 8 1.274 1.275 1.291 1.292 1.330 1.358 1.403 1.474 1.551 1.656 1.787 1.951 7 1.255 1.245 1.256 1.271 1.284 1.326 1.373 1.435 1.509 1.625 1.746 1.911 6 1.213 1.211 1.214 1.232 1.254 1.292 1.339 1.408 1.491 1.606 1.744 1.928 5 1.135 1.145 1.145 1.158 1.176 1.204 1.245 1.308 1.384 1.496 1.630 1.810 4 1.073 1.071 1.076 1.082 1.096 1.138 . 1.176 1.230 1.311 1.419 1.566 1.754 3 0.970 0.956 0.963 0.978 0.991 1.022 1.056 1.105 1.178 1.283 1.412 1.589 2 0.874 0.874 0.879 0.885 0.896 0.927 0.963 1.010 1.087 1.182 1.311 1.496 1 1.061 1.071 1.065 1.066 1.074 1.082 1.115 1.145 1.191 1.262 1.364 1.504 Plate 0.886 0.880 0.877 0.878 0.884 0.893 0.911 0.933 0.964 1.008 1.063 1.137 Axial 1.410 1.421 1.444 1.444 1.475 1.492 1.511 1.549 1.578 1.611 1.649 1.684 A-13

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-13. MURR fuel element 5 peaking factor for the extreme burnup c o r e -

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.932 0.769 0.677 0.615 0.577 0.550 0.524 0.504 0.483 0.465 0.461 0.449 23 0.960 0.760 0.647 0.574 0.517 0.493 0.459 0.439 0.415 0.409 0.398 0.389 22 1.093 0.889 0.758 0.662 0.609 0.573 0.534 0.503 0.491 0.475 0.470 0.458 21 1.259 1.014 0.868 0.779 0.709 0.652 0.624 0.598 0.575 0.561 0.547 0.533 20 1.420 1.147 0.981 0.868 0.814 0.756 0.724 0.690 0.668 0.649 0.624 0.621 19 1.590 1.285 1.102 0.997 0.927 0.873 0.828 0.783 0.763 0.741 0.723 0.706 18 1.806 1.448 1.253 1.126 1.039 0.975 0.932 0.897 0.861 0.827 0.819 0.804 17 2.018 1.642 1.422 1.262 1.163 1.088 1.027 0.991 0.958 0.935 0.912 0.900 16 2.241 1.832 1.589 1.406 1.300 1.203 1.150 1.108 1.070 1.045 1.023 0.995 15 2.462 1.999 1.725 1.550 1.421 1.331 1.269 1.221 1.187 1.158 1.125 1.102 14 2.657 2.171 1.870 1.669 1.537 1.450 1.354 1.313 1.271 1.242 1.222 1.210 13 2.824 2.297 1.984 1.784 1.649 1.550 1.466 1.410 1.374 1.342 1.321 1.291 12 2.951 2.399 2.085 1.862 1.708 1.606 1.534 1.470 1.436 1.412 1.387 1.374 11 3.082 2.505 2.178 1.948 1.780 1.673 1.598 1.532 1.487 1.445 1.423 1.417 10 3.173 2.563 2.221 1.992 1.830 1.716 1.634 1.572 1.525 1.491 1.480 1.466 9 3.222 2.608 2.236 2.010 1.855 1.734 1.660 1.600 1.538 1.513 1.489 1.460 8 3.231 2.619 2.253 2.018 1.845 1.720 1.651 1.583 1.526 1.499 1.482 1.467 7 3.192 2.581 2.221 1.976 1.819 1.699 1.609 1.562 1.513 1.486 1.453 1.426 6 3.081 2.487 2.145 1.895 1.749 1.631 1.552 1.491 1.444 1.415 1.394 1.380 5 2.885 2.322 1.999 1.783 1.650 1.540 1.446 1.402 1.352 1.325 1.298 1.280 4 2.675 2.162 1.843 1.650 1.505 1.407 1.339 1.289 1.259 1.228 1.209 1.188 3 2.400 1.968 1.679 1.506 1.364 1.283 1.211 1.164 1.126 1.101 1.081 1.068 2 2.195 1.773 1.510 1.345 1.233 1.147 1.089 1.051 1.014 0.992 0.982 0.962 1 2.256 1.882 1.665 1.524 1.437 1.369 1.326 1.281 1.247 1.233 1.206 1.193 Plate 1.873 1.520 1.310 1.172 1.079 1.011 0.961 0.925 0.895 0.875 0.860 0.847 Axial 1.395 1.393 1.390 1.392 1.390 1.386 1.396 1.399 1.389 1.397 1.399 1.401 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.439 0.432 0.418 0.419 0.413 0.410 0.410 0.401 0.405 0.412 0.437 0.484 23 0.383 0.375 0.368 0.361 0.359 0.363 0.366 0.372 0.378 0.396 0.433 0.495 22 0.456 0.446 0.440 0.438 0.433 0.431 0.437 0.440 0.447 0.474 0.507 0.570 21 0.530 0.527 0.518 0.512 0.513 0.498 0.510 0.514 0.527 0.547 0.590 0.666 20 0.611 0.601 0.593 0.590 0.587 0.580 0.592 0.596 0.616 0.643 0.688 0.774 19 0.695 0.684 0.671 0.677 0.672 0.672 0.680 0.686 0.707 0.734 0.802 0.896 18 0.796 0.782 0.772 0.767 0.759 0.761 0.770 0.781 0.803 0.838 0.906 1.014 17 0.893 0.883 0.867 0.857 0.865 0.863 0.866 0.876 0.914 0.944 1.017 1.147 16 0.993 0.977 0.961 0.973 0.970 0.962 0.973 0.992 1.018 1.069 1.153 1.312 15 1.091 1.084 1.077 1.077 1.080 . 1.084 1.100 1.135 1.167 1.238 1.354 1.559 14 1.196 1.187 1.185 1.180 1.189 1.204 1.232 1.263 1.313 1.414 1.550 1.813 13 1.285 1.271 1.272 1.276 1.275 1.304 1.338 1.377 1.450 1.546 1.713 1.989 12 1.373 1.354 1.342 1.360 1.358 1.372 1.408 1.453 1.543 1.644 1.826 2.144 11 1.415 1.406 1.393 1.393 1.415 1.442 1.481 1.533 1.623 1.739 1.937 2.268 10 1.445 1.441 1.430 1.439 1.461 1.481 1.521 1.586 1.667 1.802 2.032 2.411 9 1.464 1.450 1.456 1.459 1.476 1.498 1.555 1.617 1.726 1.884 2.121 2.553 8 1.436 1.441 1.452 1.457 1.470 1.509 1.562 1.631 1.738 1.911 2.173 2.629 7 1.412 1.414 1.408 1.426 1.454 1.483 1.542 1.614 1.730 1.901 2.165 2.636 6 1.355 1.366 1.360 1.367 1.392 1.416 1.472 1.563 1.675 1.837 2.099 2.551 5 1.283 1.283 1.284 1.296 1.322 1.341 1.394 1.471 1.585 1.736 1.984 2.405 4 1.181 1.179 1.180 1.193 1.213 1.236 1.280 1.347 1.458 1.610 1.848 2.251 3 1.060 1.060 1.060 1.073 1.088 1.124 1.160 1.230 1.318 1.462 1.677 2.054 2 0.955 0.951 0.957 0.964 0.985 1.014 1.055 1.111 1.196 1.326 1.531 1.875 1 1.192 1.187 1.182 1.186 1.196 1.216 1.243 1284 1.345 1.445 1.608 1.913 Plate 0.840 0.835 0.830 0.833 0.840 0.851 0.874 0.905 0.955 1.029 1.150 1.361 Axial 1.409 1.404 1.418 1.416 1.420 1.434 1.444 1.457 1.471 1.501 1.527 1.566 A-14

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-14. MURR fuel element 6 peaking factor for the extreme burnup c<>re -

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.748 0.658 0.586 0.544 0.507 0.483 0.459 0.443 0.434 0.420 0.404 0.395 23 0.777 0.659 0.578 0.512 0.478 0.448 0.416 0.394 0.377 0.361 0.348 0.334 22 0.852 0.732 0.655 0.600 0.534 0.498 0.478 0.448 0.442 0.426 0.415 0.393 21 0.998 0.849 0.768 0.690 0.636 0.594 0.557 0.541 0.516 0.496 0.483 0.472 20 1.065 0.923 0.834 0.758 0.704 0.669 0.634 0.611 0.587 0.572 0.554 0.540 19 1.199 1.061 0.949 0.872 0.803 0.763 0.728 0.697 0.672 0.647 0.636 0.609 18 1.317 1.163 1.043 0.955 0.886 0.841 0.809 0.770 0.740 0.724 0.709 0.690 17 1.468 1.299 1.180 1.082 1.019 0.950 0.901 0.863 0.836 0.814 0.792 0.770 16 1.587 1.411 1.282 1.183 1.109 1.052 0.994 0.972 0.928 0.895 0.869 0.851 15 1.739 1.560 1.423 1.311 1.217 1.148 1.103 1.061 1.027 1.001 0.978 0.969 14 1.860 1.645 1.520 1.406 1.305 1.247 1.193 1.138 1.119 1.091 1.079 1.062 13 1.944 1.753 1.607 1.492 1.405 1.324 1.281 1.232 1.203 1.179 1.152 1.136 12 2.020 1.817 1.678 1.561 1.463 1.385 1.334 1.287 1.260 1.235 1.222 1.202 11 2.098 1.880 1.730 1.621 1.533 1.452 1.397 1.341 1.311 1.281 1.263 1.258 10 2.176 1.957 1.802 1.673 1.559 1.493 1.429 1.388 1.350 1.313 1.305 1.294 9 2.234 2.016 1.840 1.692 1.598 1.499 1.447 1.390 1.364 1.341 1.327 1.312 8 2.285 2.045 1.865 1.724 1.610 1.517 1.458 1.411 1.381 1.344 1.330 1.314 7 2.251 2.005 1.817 1.684 1.572 1.496 1.434 1.374 1.343 1.318 1.305 1.286 6 2.262 2.003 1.820 1.662 1.543 1.450 1.388 1.359 1.310 1.276 1.260 1.247 5 2.119 1.887 1.694 1.564 1.456 1.383 1.310 1.266 1.233 1.198 1.179 1.172 4 2.035 1.774 1.606 1.476 1.366 1.280 1.217 1.184 1.158 1.133 1.111 1.094 3 1.873 1.618 1.454 1.327 1.226 1.160 1.105 1.077 1.039 1.022 1.001 0.984 2 1.697 1.496 1.318 1.195 1.116 1.055 1.000 0.958 0.932 0.917 0.897 0.887 1 1.740 1.560 1.426 1.348 1.271 1.228 1.199 1.173 1.140 1.123 1.098 1.095 Plate 1.598 1.417 1.287 1.186 1.106 1.046 1.001 0.966 0.939 0.916 0.900 0.886 Axial 1.359 1.372 1.379 1.383 1.384 1.378 1.385 1.390 1.398 1.394 1.405 1.410 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.376 0.365 0.359 0.346 0.339 0.323 0.309 0.294 0.280 0.265 0.247 0.235 23 0.327 0.321 0.312 0.303 0.299 0.288 0.281 0.273 0.264 0.247 0.239 0.231 22 0.385 0.372 0.363 0.357 0.354 0.338 0.323 0.315 0.303 0.286 0.270 0.254 21 0.453 0.437 0.432 0.422 0.411 0.399 0.388 0.376 0.362 0.348 0.329 0.308 20 0.526 0.515 0.496 0.483 0.474 0.460 0.445 0.429 0.407 0.387 0.367 0.341 19 0.597 0.587 0.575 0.557 0.545 0.534 0.513 . 0.494 0.476 0.451 0.420 0.393 18 0.675 0.657 . 0.643 0.626 0.601 0.588 0.563 0.543 0.520 0.492 0.466 0.426 17 0.758 0.750 0.726 0.703 0.680 0.667 0.652 0.627 0.600 0.570 0.536 0.505 16 0.840 0.821 0.815 0.794 0.783 0.759 0.748 0.732 0.701 0.677 0.648 0.611 15 0.941 0.931 0.917 0.919 0.912 0.916 0.911 0.915 0.922 0.936 0.949 0.980 14 1.047 1.038 1.032 1.036 1.045 1.059 1.075 1.112 1.150 1.199 1.274 1.384 13 1.126 1.120 1.124 1.139 1.148 1.162 1.201 1.244 1.293 1.377 1.479 1.626 12 1.186 1.193 1.189 1.201 1.214 1.240 1.285 1.335 1.400 1.489 1.609 1.768 11 1.236 1.241 1.250 1.256 1.278 1.311 1.361 1.415 1.491 1.585 1.735 1.906 10 1.284 1.285 1.296 1.312 1.328 1.362 1.402 1.462 1.548 1.669 1.830 2.046 9 1.304 1.298 1.310 1.320 1.344 1.375 1.423 1.493 1.589 1.699 1.862 2.089 8 1.309 1.302 1.298 1.322 1.339 1.371 1.422 1.496 1.589 1.727 1.894 2.125 7 1.265 1.271 1.279 1.300 1.317 1.357 1.387 1.458 1.556 1.682 1.843 2.075 6 1.251 1.236 1.242 1.255 1.288 1.309 1.359 1.428 1.526 1.649 1.821 2.062 5 1.169 1.164 1.164 1.180 1.199 1.237 1.283 1.348 1.429 1.560 1.718 1.941 4 1.087 1.084 1.098 1.109 1.123 1.150 1.201 1.259 1.349 1.466 1.635 1.878 3 0.976 0.972 0.981 0.987 1.005 1.039 1.077 1.133 1.218 1.341 1.500 1.715 2 0.881 0.882 0.887 0.903 0.918 0.941 0.980 1.042 1.116 1.231 1.383 1.597 1 1.093 1.093 1.093 1.099 1.097 1.122 1.149 1.182 1.246 1.329 1.445 1.615 Plate 0.875 0.869 0.867 0.869 0.873 0.884 0.901 0.927 0.964 1.017 1.089 1.193 Axial 1.422 1.424 1.437 1.447 1.464 1.479 1.502 1.534 1.567 1.615. 1.653 1.694 A-15

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-15. MURR fuel element 7 peaking factor for the extreme burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.789 0.669 0.600 0.546 0.500 0.471 0.452 0.432 0.414 0.410 0.394 0.383 23 0.830 0.676 0.578 0.511 0.461 0.426 0.397 0.375 0.361 0.348 0.337 0.328 22 0.928 0.759 0.654 0.582 0.539 0.495 0.464 0.437 0.415 0.403 0.390 0.381 21 1.058 0.879 0.757 0.670 0.624 0.574 0.539 0.520 0.494 0.480 0.461 0.451 20 1.178 0.974 0.855 0.763 0.696 0.661 0.624 0.589 0.565 0.542 0.533 0.514 19 1.325 1.109 0.971 0.878 0.808 0.755 0.710 0.673 0.648 0.620 0.592 0.584 18 1.482 1.251 1.094 0.996 0.909 0.841 0.792 0.758 0.724 0.696 0.674 0.660 17 1.649 1.409 1.222 1.102 1.016 0.943 0.894 0.841 0.807 0.784 0.760 0.733 16 1.809 1.516 1.347 1.200 1.097 1.025 0.972 0.922 0.888 0.859 0.838 0.819 15 1.961 1.651 1.466 1.307 1.204 1.113 1.067 1.009 0.975 0.942 0.917 0.889 14 2.106 1.802 1.579 1.422 1.298 1.211 1.141 1.092 1.054 1.021 0.993 0.977 13 2.239 1.892 1.665 1.514 1.396 1.291 1.217 1.170 1.133 1.094 1.064 1.045 12 2.322 1.985 1.749 1.590 1.450 1.355 1.285 1.229 1.198 1.166 1.136 1.113 11 2.398 2.059 1.815 1.643 1.514 1.418 1.353 1.293 1.248 1.212 1.179 1.165 10 2.494 2.134 1.880 1.696 1.571 1.476 1.393 1.351 1.305 1.254 1.233 1.217 9 2.557 2.166 1.907 1.721 1.588 1.496 1.433 1.381 1.334 1.292 1.269 1.247 8 2.612 2.221 1.939 1.771 1.623 1.530 1.446 1.386 1.346 1.309 1.285 1.245 7 2.591 2.180 1.919 1.725 1.583 1.483 1.436 1.381 1.323 1.285 1.258 1.242 6 2.552 2.143 1.866 1.699 1.571 1.464 1.388 1.328 1.294 1.249 1.224 1.204 5 2.400 2.037 1.784 1.609 1.486 1.387 1.311 1.251 1.207 1.182 1.159 1.143 4 2.264 1.893 1.662 1.500 1.383 1.298 1.225 1.172 1.145 1.116 1.082 1.069 3 2.054 1.736 1.514 1.356 1.238 1.160 1.101 1.059 1.019 0.992 0.975 0.968 2 1.883 1.573 1.359 1.225 1.128 1.057 0.999 0.955 0.924 0.910 0.890 0.875 1 1.914 1.668 1.508 1.405 1.317 1.260 1.205 1.173 1.141 1.115 1.107 1.084 Plate 1.866 1.578 1.385 1.251 1.151 1.077 1.021 0.978 0.944 0.916 0.894 0.877 Axial 1.381 1.389 1.381 1.396 1.391 1.402 1.397 1.399 1.407 1.410 1.418 1.403 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.370 0.360 0.352 0.344 0.329 0.319 0.311 0.295 0.287 0.277 0.272 0.270 23 0.320 0.307 0.303 0.294 0.288 0.280 0.272 0.268 0.262 0.263 0.263 0.276 22 0.367 0.361 0.348 0.342 0.338 0.328 0.321 0.314 0.311 0.304 0.298 0.309 21 0.435 0.423 0.411 0.407 0.396 0.385 0.373 0.366 0.360 0.359 0.349 0.359 20 0.494 0.480 0.469 0.458 0.447 0.444 0.433 0.422 0.414 0.407 0.402 0.405 19 0.569 0.555 0.538 0.523 0.515 0.503 0.488 0.479 0.474 0.468 0.462 0.465 18 0.640 0.622 0.607 0.591 0.580 0.565 0.554 0.541 0.525 0.524 0.510 0.514 17 0.708 0.696 0.678 0.666 0.652 0.636 0.626 0.614 0.604 0.587 0.581 0.587 16 0.790 0.774 0.761 0.743 0.722 0.715 0.701 0.692 0.678 0.676 0.670 0.679 15 0.880 0.865 0.849 0.841 0.828 0.820 0.810 0.815 0.823 0.836 0.856 0.910 14 0.959 0.953 0.936 0.931 0.925 0.915 0.917 0.938 0.962 1.000 1.071 1.178 13 1.032 1.018 1.015 1.012 1.017 1.015 1.019 1.030 1.079 1.116 1.204 1.328 12 1.093 1.080 1.071 1.070 1.065 1.073 1.086 1.109 1.148 1.212 1.302 1.454 11 1.151 1.134 1.135 1.138 1.139 1.146 1.161 1.198 1.238 1.315 1.420 1.591 10 1.200 . 1.196 1.191 1.183 1.186 1.200 1.222 1.273 1.349 1.442 1.570 1.793 9 1.237 1.223 1.219 1.225 1.231 1.253 1.292 1.339 1.425 1.533 1.696 1.956 8 1.241 1.239 1.232 1.242 1.265 1.285 1.318 1.367 1.451 1.590 1.758 2.036 7 1.240 1.227 1.224 1.226 1.245 1.277 1.314 1.366 1.453 1.576 1.746 2.026 6 1.193 1.190 1.184 1.193 1.208 1.238 1.274 1.326 1.414 1.546 1.730 2.014 5 1.137 1.129 1.125 1.135 1.152 1.174 1.212 1.261 1.356 1.458 1.653 1.916 4 1.060 1.058 1.052 1.064 1.077 1.102 1.140 1.201 1.273 1.382 1.559 1.832 3 0.951 0.958 0.955 0.958 0.971 1.005 1.042 1.080 1.160 1.262 1.428 1.692 2 0.867 0.851 0.868 0.864 0.891 0.910 0.946 0.988 1.065 1.163 1.325 1.586 1 1.073 1.074 1.075 1.072 1.081 1.097 1.118 1.154 1.206 1.294 1.425 1.642 Plate 0.864 0.854 0.847 0.844 0.845 0.850 0.861 0.881 0.918 0.970 1.050 1.185 Axial 1.418 1.432 1.435 1.452 1.478 1.491 1.510 1.530 1.563 1.618 1.651 1.695 A-16

ATTACHMENT6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table A-16. MURR fuel element 8 peaking factor for the extreme burnup core - )

Axial Fuel plate number node 1 2 3 4 5 6 7 8 9 10 11 12 24 0.682 0.601 *0.546 0.504 0.470 0.435 0.420 0.398 0.389 0.368 0.359 0.347 23 0.695 0.594 0.518 0.471 0.431 0.393 0.375 0.362 0.339 0.322 0.316 0.298 22 0.768 0.680 0.597 0.540 0.492 0.455 0.433 0.413 0.391 0.377 0.365 0.350 21 0.892 0.778 0.697 0.628 0.575 0.523 0.493 0.476 0.454 0.438 0.429 0.409 20 0.956 0.846 0.765 0.696 0.641 0.602 0.565 0.536 0.515 0.493 0.474 0.460 19 1.092 0.966 0.868 0.796 0.731 0.681 0.640 0.603 0.581 0.566 0.539 0.519 18 1.152 1.051 0.944 0.856 0.797 0.734 0.689 0.657 0.637 0.616 0.596 0.568 17 1.266 1.151 1.057 0.960 0.883 0.827 0.780 0.741 0.708 0.681 0.665 0.636 16 1.338 1.216 1.112 1.033 0.952 0.892 0.842 0.801 0.772 0.740 0.716 0.690 15 1.455 1.327 1.209 1.123 1.038 0.966 0.921 0.879 0.846 0.799 0.777 0.752 14 1.524 1.401 1.292 1.184 1.114 1.030 0.981 0.934 0.898 0.854 0.823 0.797 13 1.622 1.493 1.366 1.258 1.183 1.107 1.053 0.994 0.960 0.918 0.885 0.858 12 1.677 1.546 1.422 1.310 1.227 1.149 1.095 1.046 1.008 0.969 0.937 0.903 11 1.739 1.603 1.483 1.371 1.291 1.222 1.150 1.104 1.062 1.023 0.993 0.965 10 1.833 1.699 1.570 1.434 1.345 1.275 1.215 1.158 1.118 1.082 1.061 1.033 9 1.893 1.740 1.604 1.475 1.390 1.316 1.253 1.200 1.172 1.132 1.101 1.080 8 1.977 1.806 1.666 1.534 1.430 1.349 1.282 1.238 1.199 1.182 1.155 1.135 7 1.958 1.789 1.645 1.519 1.425 1.347 1.286 1.233 1.199 1.178 1.146 1.133 6 2.002 1.812 1.652 1.523 1.415 1.330 1.273 1.206 1.173 1.151 1.134 1.116 5 1.894 1.715 1.559 1.438 1.336 1.262 1.199 1.156 1.119 1.095 1.063 1.062 4 1.856 1.659 1.506 1.364 1.273 1.199 1.134 1.097 1.056 1.038 1.020 1.004 3 1.674 1.485 1.339 1.239 1.145 1.087 1.034 0.986 0.959 0.937 0.909 0.907 2 1.586 1.396 1.247 1.131 1.048 0.991 0.936 0.911 0.877 0.852 0.835 0.824 1 1.633 1.486 1.367 1.280 1.232 1.161 1.131 1.093 1.069 1.056 1.036 1.027 Plate 1.739 1.575 1.436 1.319 1.230 1.154 1.097 1.050 1.014 0.983 0.956 0.934 Axial 1.366 1.366 1.377 1.380 1.380 1.387 1.392 1.400 1.404 1.428 1.433 1.443 Axial Fuel plate number node 13 14 15 16 17 18 19 20 21 22 23 24 24 0.340 0.327 0.317 0.307 0.294 0.281 0.269 0.261 0.241 0.223 0.208 0.194 23 0.283 0.279 0.275 0.265 0.256 0.248 0.239 0.230 0.220 0.209 0.200 0.188 22 0.339 0.326 0.316 0.306 0.296 0.280 0.276 0.261 0.249 0.234 0.223 0.202 21 0.397 0.380 0.367 0.365 0.346 0.329 0.320 0.308 0.292 0.274 0.254 0.233 20 0.442 0.428 0.416 0.403 0.383 0.370 0.356 0.336 0.318 0.297 0.268 0.247 19 0.502 0.483 0.469 0.460 0.439 0.421 0.406 0.390 0.365 0.345 0.318 0.284 18 0.555 0.540 0.519 0.496 0.482 0.465 0.445 0.420 0.397 0.369 0.334 0.295 17 0.612 0.594 0.577 0.558 0.545 0.514 0.498 0.469 0.436 0.410 0.375 0.330 16 0.667 0.645 0.627 0.610 0.587 0.562 0.537 0.509 0.476 0.439 0.396 0.352 15 0.728 0.701 0.676 0.661 0.637 0.615 0.585 0.555 0.521 0.483 0.434 0.383 14 0.775 0.760 0.722 0.698 0.681 0.648 0.622 0.593 0.548 0.507 0.456 0.406 13 0.834 0.801 0.776 0.753 0.724 0.698 0.670 0.635 0.597 0.558 0.504 0.445 12 0.878 0.856 0.831 0.803 0.777 0.748 0.715 0.683 0.641 0.596 0.540 0.483 11 0.939 0.916 0.892 0.871 0.843 0.827 0.798 0.767 0.734 0.688 0.637 0.579 10 1.011 0.995 0.984 0.962 0.951 0.942 0.938 0.933 0.923 0.921 0.916 0.910 9 1.060 1.057 1.056 1.049 1.051 1.066 1.080 1.111 1.145 1.196 1.257 1.323 8 1.118 1.100 1.102 1.111 1.122 1.145 1.173 1.214 1.275 1.348 1.441 1.546 7 1.105 1.107 1.112 1.111 1.123 1.150 1.185 1.230 1.297 1.377 1.479 1.605 6 1.098 1.097 1.097 1.113 1.124 1.162 1.191 1.244 1.310 1.397 1.517 1.677 5 1.043 1.041 1.037 1.060 1.077 1.098 1.141 1.191 1.251 1.348 1.466 1.616 4 0.993 0.986 0.985 0.998 1.016 1.034 1.084 1.133 1.205 1.297 1.428 1.597 3 0.899 0.908 0.902 0.911 0.921 0.944 0.992 1.028 1.095 1.187 1.304 1.459 2 0.817 0.816 0.815 0.822 0.836 0.860 0.900 0.947 1.011 1.110 1.227 1.401 1 1.018 1.015 1.012 1.009 1.017 1.036 1.056 1.092 1.141 1.203 1.298 1.434 Plate 0.913 0.898 0.885 0.876 0.867 0.863 0.865 0.868 0.875 0.891 0.914 0.949 Axial 1.454 1.464 1.492 1.509 1.538 1.598 1.635 1.702 1.777 1.861 1.970 2.097 A-17

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441R00031/B APPENDIX B TARGET ASSEMBLY LINEAR POWER B-1

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Table B-1. Target assembly linear power for the maximum burnup core - )

• I I I I I I I I I B-2

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table 8-2. Target assembly linear power for the average burnup core -

B-3

~-----------------------------

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031/B Table B-3. Target assembly linear power for the minimum burnup core - )

• I I

ATTACHMENT 6 M0-99 Target Assembly Nuclear Design for Once-Through Operation 30441 R00031 /B Table B-4. Target assembly linear power for the extreme burnup core -

• I I I I I 8-5

ATTACHMENT 6

+GENERAL ATOMleS P.O. BOX 85608 SAN DIEGO, CA 92186-5608 (858) 455-3000