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Copyright by Alexander Gary Fay 2014 the Dissertation Committee for Alexander Gary Fay Certifies That This Is the Approved Version of the Following Dissertation Copyright by Alexander Gary Fay 2014 The Dissertation Committee for Alexander Gary Fay certifies that this is the approved version of the following dissertation: Characterization of Sources of Radioargon in a Research Reactor Committee: Steven Biegalski, Supervisor Derek Haas Sheldon Landsberger Joshua Peterson Erich Schneider Characterization of Sources of Radioargon in a Research Reactor by Alexander Gary Fay, B.S.; M.S.E. Dissertation Presented to the Faculty of the Graduate School of The University of Texas at Austin in Partial Fulfillment of the Requirements for the Degree of Doctor of Philosophy The University of Texas at Austin May 2014 Dedication For my family. Acknowledgements The author would like to acknowledge Dr. Steven Biegalski for his invaluable guidance and support throughout this project, the members of the Dissertation Committee for their time and input, and the staff of the Nuclear Engineering Teach- ing Laboratory at the University of Texas at Austin. The author would like to acknowledge the assistance of former University of Texas at Austin colleagues who all made tangible contributions to the progress of this research: Kenneth Dayman, James Armstrong, Robert Flanagan, Franziska Klingberg, Joseph Graham, and Christopher van der Hoeven. The author would also like to acknowledge the persistent support of colleagues in the Office of Nonproliferation Research and Development at the National Nuclear Security Administration. The author gratefully acknowledges the NOAA Air Resources Laboratory for the provision of the HYSPLIT transport and dispersion model, along with Los Alamos National Laboratory for the provision of the Monte Carlo N-Particle radiation trans- port model, both of which were used in this publication. This research was performed, in part, under appointment to the U.S. Department of Energy Nuclear Nonproliferation International Safeguards Graduate Fellowship Program sponsored by the National Nuclear Security Administration’s Office of Non- proliferation and International Security. v Characterization of Sources of Radioargon in a Research Reactor Alexander Gary Fay, Ph.D. The University of Texas at Austin, 2014 Supervisor: Steven Biegalski On Site Inspection is the final measure for verifying compliance of Member States with the Comprehensive Nuclear-Test-Ban Treaty. In order to enable the use of 37Ar as a radiotracer for On Site Inspection, the sources of radioargon background must be characterized and quantified. A radiation transport model of the University of Texas at Austin Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor was developed to simulate the neutron flux in various regions of the reactor. An activation and depletion code was written to calculate production of 37Ar in the facility based on the results of the radiation transport model. Results showed 37Ar production rates of (6.567±0.31)×102 Bq·kWh−1 in the re- actor pool and the air-filled irradiation facilities, and (5.811±0.40)×104 Bq·kWh−1 in the biological shield. Although 40Ca activation in the biological shield was found to dominate the total radioargon inventory, the contribution to the effluent release rate would be diminished by the immobility of Ar generated in the concrete matrix vi and the long diffusion path of mobile radioargon. Diffusion of radioargon out of the reactor pool was found to limit the release rate but would not significantly affect the integrated release activity. The integrated 37Ar release for an 8 hour operation at 950 kW was calculated to be (1.05±0.8)×107 Bq, with pool emissions continuing for days and biological shield emissions continuing for tens of days following the operation. Sensitivity analyses showed that estimates for the time-dependent concentra- tions of 37Ar in the NETL TRIGA could be made with the calculated buildup co- efficients or through analytical solution of the activation equations for only (n; γ) reactions in stable argon or (n; α) reactions in 40Ca. Analyses also indicated that, for a generalized system, the integrated thermal flux can be used to calculate the buildup due to air activation and the integrated fast flux can be used to calculate the buildup due to calcium activation. Based on the results of the NETL TRIGA, an estimate of the global research reactor source term for 37Ar and an estimate of ground-level 37Ar concentrations near a facility were produced. vii Table of Contents Table of Contents viii List of Figures x List of Figures . xi List of Tables xii List of Tables . xiii 1 Introduction 1 1.1 Scope of the Research . 7 1.2 Overview . 8 2 Literature Review 9 2.1 On Site Inspection . 9 2.2 Argon as a Reactor Effluent . 10 2.3 Natural Sources of Radioargon Background . 12 2.4 Radioargon Production and Measurement . 13 3 Methodology 18 3.1 Radioargon Basics . 18 3.2 Radiation Transport . 25 3.3 Air Content of Water . 38 3.4 Diffusion for Ca-Generated Argon . 41 3.5 Atmospheric Transport . 44 3.6 Experiments . 48 4 Model and Code Validation 53 4.1 Reactor Model Validation . 53 4.2 Depletion Code Validation . 60 5 Results 64 5.1 Uncertainty . 64 5.2 In-Core Sources . 68 5.3 Ex-Core Sources . 77 viii 5.4 Other Sources . 88 5.5 Diffusion and Tranport . 92 5.6 Atmospheric Transport . 98 5.7 41Ar Benchmarking . 102 6 Generalization of Results 104 6.1 Thermal and Fast Flux Dependence . 104 6.2 Proxy Isotope Estimates . 109 6.3 Global Research Reactor Fleet . 112 7 Sensitivity Analyses 116 7.1 Operational Cycles . 117 7.2 Epithermal and Fast Neutrons . 120 7.3 Depletion Code Fidelity . 122 8 Summary and Discussion 126 8.1 Summary of Results . 126 8.2 Conclusions . 131 8.3 Future Work . 132 Appendix A MCNPX - Reactor Model 136 Appendix B Data Processing 161 Appendix C Research Reactor Utilization 180 References 187 ix List of Figures 3.1 Cross-section for 36Ar(n,γ)37Ar, 38Ar(n,γ)39Ar, and 40Ar(n,γ)41Ar . 22 3.2 Cross-section for 40Ca(n,α)37Ar . 24 3.3 Vertical cross-section view of the MCNPX TRIGA core model with control rods inserted . 26 3.4 Lateral cross-section view of the MCNPX TRIGA core with control rods inserted . 27 3.5 Lateral cross-section view of the complete TRIGA reactor modeled in MCNPX . 28 4.1 Experimental and modeled flux of the RSR facility at 950 kW . 57 5.1 Uncertainty associated with the thermal cross-sections for the reac- tions 36Ar(n,γ)37Ar, 38Ar(n,γ)39Ar, and 40Ar(n,γ)41Ar . 67 5.2 Buildup of 37Ar, 39Ar, and 41Ar in the 3L facility at 950 kW . 69 5.3 Group flux of the 3L facility at 950 kW . 70 5.4 Buildup of 37Ar, 39Ar, and 41Ar in the RSR facility at 950 kW . 73 5.5 Flux of the PTS facility at 950 kW . 76 5.6 Radial segments of the reactor pool used for cell-based flux tallies . 79 5.7 Buildup of 37Ar, 39Ar, and 41Ar in the reactor pool at 950 kW . 80 5.8 Buildup of 37Ar, 39Ar, and 41Ar in the biological shield at 950 kW . 85 5.9 Tallying regions for calculating radioargon buildup in the biological shield concrete around the beamports . 87 5.10 Maximum buildup of 37Ar, 39Ar, and 41Ar in fuel element B2 during operation at 950 kW. 91 5.11 Vented fraction of 37Ar from the biological shield following a 8 h irradiation at 950 kW . 94 5.12 Vented activity (blue) and exhalation rate (red) of 37Ar from the bio- logical shield following a 8 h irradiation at 950 kW . 95 5.13 Release rate of 37Ar as a function of time from the beamports and RSR, the reactor pool, and the biological shield for an 8 h operation at 950 kW . 96 5.14 Ground level concentration of 37Ar for a simulated operation of the NETL TRIGA for 8 h at 950 kW . 99 x 6.1 Volume-normalized buildup constants as a function of thermal flux for the air-filled irradiation facilities . 105 6.2 Volume-normalized buildup constants as a function of thermal flux for the tally regions of the reactor pool . 107 6.3 Volume-normalized buildup constants as a function of fast flux for the tally regions of the reactor biological shield . 108 6.4 Distribution of research reactor utilization, according to the IAEA Re- search Reactor Database . 113 7.1 Difference in calculated 37Ar in the 3L for the buildup coefficient ver- sus the depletion code . 118 7.2 Buildup of 37Ar in a sample of concrete as calculated using different nuclear data sets for the 40Ca(n; α) reaction . 125 xi List of Tables 3.1 Decay data for the primary radioargon isotopes . 19 3.2 Mass fraction composition of SRM 1888a, Portland Cement . 51 4.1 Results of 3L irradiation of Al/Co flux wires . 54 4.2 Results of RSR irradiation of Al/Co flux wires . 55 4.3 Experimental and MCNPX-generated flux at the end of each beamport 59 4.4 Activity of the activation products in the SRM1888a sample . 60 4.5 Comparison of air activation results from the depletion code and ORIGEN-ARP . 62 4.6 Uncertainty in depletion code results due to computational variance . 62 5.1 Uncertainty in37Ar buildup for sample depletion code results . 66 5.2 Activity of the radioargon isotopes in the 3L during an 8 h irradiation at 950 kW . 71 5.3 Buildup constants for radioargon in the 3L . 71 5.4 Activity of the radioargon isotopes in the RSR during an 8 h irradia- tion at 950 kW . 72 5.5 Buildup constants for radioargon in the RSR . 72 5.6 Activity of the radioargon isotopes in the PTS during an 8 h irradia- tion at 950 kW .
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