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Introduction MONTE CARLO NEUTRONIC SIMULATIONS FOR A NEW APPROACH TO PARAMETRIC INAA AND MO-99 PRODUCTION FEASIBILITY AT MURR _______________________________________ A Dissertation presented to the Faculty of the Graduate School at the University of Missouri-Columbia _______________________________________________________ In Partial Fulfillment of the Requirements for the Degree Doctor of Philosophy _____________________________________________________ by NICKIE J.J. PETERS Dr. John David Robertson, Dissertation Supervisor August 2009 ACKNOWLEDGMENTS I would like to thank Professor J. David Robertson for the great opportunity to join his research group and for mentoring me over the past years, my advisory committee, Professors William .H. Miller, Silvia S. Jurisson and Carol A. Deakyne for steady support, advice and guidance. Dr. John D. Brockman, Dr. Kiratadas Kutikkad, and Mr. Charles McKibben are thanked for being excellent and intelligent resources, the University of Missouri Chemistry Department and the University of Missouri Research Reactor facility are thanked for educational and research opportunities and support. ii ABSTRACT A novel approach to parametric instrumental neutron activation analysis at MURR has been established. In particular, a detailed MCNP5 steady-state model of the MURR core was developed. The model, which was based on the most recent continuous-energy neutron data from the ENDF and JEFF libraries, was used to compute the local continuous-energy neutron flux distribution. By coupling the computed flux spectrum to the energy-dependent (n, γ) cross-sections for a range of nuclides, their intrinsic reaction rates were predicted in irradiation channel ROW2. The model was initially benchmarked by measuring the intrinsic (n, γ) reaction rates for a set of mostly dilute single-element standards in ROW2. Results show that the model predicts the absolute reaction rates of many nuclides including those with high epithermal sensitivity (e.g., Au-197 and Zr-96), and non-1/v nuclides (e.g. Lu-176) within ±5% of the measured values. Using predicted (n, γ) reaction-rates characterized as the parameterπ theo , trace-elemental concentrations were determined in NIST standard reference materials, bovine liver, obsidian and coal fly ash. The agreements with the certified values were generally within ±5%. The new methodology has produced agreements with the certified values that are better for a greater number of elements than k0. The model was also combined with MONTEBURNS and ORIGEN to test the feasibility of Mo-99 production at MURR from fissioning LEU. Results from a 5-gram low-enriched uranium target show predictions of Mo-99 end-of- irradiation yields are within 3% of the measured value. This dissertation entails a complete study of the MCNP5 model and the new neutron activation analysis method. iii CONTENTS ACKNOWLEDGMENTS ......................................................................................... II ABSTRACT...............................................................................................................III LIST OF FIGURES …………………………………………….. ........................... VI LIST OF TABLES………………………………………………. ........................XIII CHAPTER 1 INTRODUCTION................................................................................1 1.1 OVERVIEW OF NEUTRON ACTIVATION ANALYSIS ...............................................................................1 1.2 NEUTRON TRANSPORT THEORY ..........................................................................................................5 1.3 FLUX MODELS FOR ACTIVATION APPLICATIONS...............................................................................14 CHAPTER 2 THEORETICAL AND EXPERIMENTAL METHODS ...............19 2.1 THEORETICAL APPROACH: THE MCNP5 MURR CORE MODEL........................................................19 2.2 MCNP5 CRITICALITY CALCULATIONS..............................................................................................33 2.3 EXPERIMENTAL APPROACH...............................................................................................................43 2.4 COUNTING AND DETECTOR EFFICIENCY CALIBRATION.....................................................................52 CHAPTER 3 MURR CORE MODEL BENCHMARKING RESULTS ..............58 3.1 MEASURED REACTION RATES IN ROW2...........................................................................................58 3.2 MODIFIED TWO-GROUP FLUX MODEL REACTION RATES IN ROW2 .................................................60 3.3 CONTINUOUS-ENERGY FLUX MODEL REACTION RATES IN ROW2...................................................69 CHAPTER 4 TRACE-ELEMENTAL CONCENTRATION ANALYSIS...........73 4.1 MASS FORMULAS FOR TRACE-ELEMENTAL CONCENTRATION ANALYSIS .........................................73 4.2 TRACE-ELEMENTAL CONCENTRATIONS IN SRM TARGETS ...............................................................77 CHAPTER 5 FLUX STABILITY STUDIES AT MURR ......................................85 5.1 FLUX STABILITY STUDIES AT MURR FOR PARAMETRIC NAA AT MURR ........................................85 5.2 FLUX STABILITY BETWEEN FUEL CYCLES .........................................................................................89 5.3 FLUX STABILITY DURING A FUEL CYCLE ........................................................................................106 CHAPTER 6 SUMMARY.......................................................................................113 CHAPTER 7 MCNP5 COUPLED ORIGEN CALCULATIONS FOR MO-99 PRODUCTION AT MURR ....................................................................................115 7.1 MOLYBDENUM-99 PRODUCTION AT MURR....................................................................................115 7.2 LEU TARGET DESIGNS....................................................................................................................118 7.3 THE MCNP5 MODELS.....................................................................................................................120 7.3.1 MCNP5 Models for Multiple LEU Target Geometries............................................................... 120 7.3.3 Static MCNP5 Flux Calculations ............................................................................................... 122 7.4 MO-99 PREDICTIONS AT MURR .....................................................................................................134 7.4.1 Benchmarking MCNP5 – ORIGEN 2 Coupled Calculations ..................................................... 134 7.4.2 LEU Target Thickness Optimization .......................................................................................... 142 7.4.3 Mo-99 Predictions for Multiple Target Arrangements............................................................... 148 7.5 CONCLUSIONS AND FUTURE WORK.................................................................................................151 APPENDIX 1 NEUTRON SELF-SHIELDING FACTOR CALCULATIONS USING MCNP5........................................................................................................155 A1.1 OVERVIEW....................................................................................................................................155 A1.2 NEUTRON SELF-SHIELDING FACTORS ..........................................................................................157 iv A1.3 MCNP5 INPUT DECK FOR SELF-SHIELDING CALCULATIONS .......................................................161 APPENDIX 2 MATERIAL DEFINITIONS FOR MURR CORE MCNP5 MODEL ....................................................................................................................164 REFERENCES: .......................................................................................................167 VITA..........................................................................................................................175 v List of Figures .................................................................... Page Fig. 1.1 A schematic of a neutron capture nuclear reaction. The delayed gamma ray is characteristic of radioactive daughter nucleus...............................................................1 Fig. 1.2 An illustration of the spatial flux distribution in neutrons ••cm−2 s − 1 from an infinite plane source through a finite, non-multiplying medium of thickness a..........10 Fig.1.3 An illustration showing the thermal and fast spatial flux distributions in neutrons ••cm−2 s − 1 for a simple critical reactor model as a function of distance from the core........................................................................................................................ 11 Fig. 1.4 A generic example of the neutron energy distribution at some distance x from the core (i.e., the source) in a non-multiplying medium using a unit lethargy scale. ..12 Fig. 1.5 A plot of the energy-dependent neutron capture cross-section for 96Zr. In the energy region below 0.5 eV depicts the thermal capture cross-section and usually varies as 1/v. The energy region above 0.5 eV (i.e., the epithermal cross-section) shows a number of resonance peaks, the largest at 338 eV. .......................................17 Fig. 2.1 A scaled cross-sectional schematic of the MURR core .................................21 Fig. 2.2 A detailed MCNP5 model of the MURR core showing the ROW2 irradiation position. The model is rotated minus 60 degrees in reference to Figure 2.1. The green area represents the graphite wedge region...................................................................23
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