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FISSION PRODUCT RELEASE from DEFECTED NUCLEAR REACTOR FUEL FXJIWINTS Ph.D. Thesis 8,3. LEWIS 1983

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FISSION PRODUCT RELEASE from DEFECTED NUCLEAR REACTOR FUEL FXJIWINTS Ph.D. Thesis 8,3. LEWIS 1983 lNiS-inl —1 2rt FISSION PRODUCT RELEASE FROM DEFECTED NUCLEAR REACTOR FUEL FXJIWINTS Ph.D. Thesis 8,3. LEWIS 1983 FISSION PRODUCT RELEASE FROM DEFECTED NUCLEAR REACTOR FUEL LCEMENTS by B..7. Lewis A thesis submitted in conformity with the requirements for the degree of Doctor of Philosophy in the University of Toronto Department of Chemical Engineering and Applied Chemistry University of Toronto B.3. Lewis, 1983 ABSTRACT The release of gaseous (krypton and xenon) arid iodine radioactive fission products from defective fuel elements is described with a semi-empirical model. The model assirrn;s pr"ecursor-_orr'.-cted "Booth diffusions! release" in the UO^ and subsequent holdup in the fuol-to-sheath j;ap. Transport in the ^ip is separately modelled with a phenoinenoloj'ical rate constant (assuming ,• please from the gap is a first order rate process), and a diffusivity constant (assuming transport in the gap is dominated by a diffusional process). Measured release data from elements possessing various states of defection are used in this analysis. One element (irradiated in an earlier experiment by MacDonald) was defected with a small dril'ed hole. A second element was machined with ?'i slits while a third element (fabricated with a porous end plug) displayed through-wall sheath hydriding. Co np.inson of measured release data with calculated values from th" model yields estimates of empirical diffusion -10 -9 coefficients for the radioactive species in the l!.'.\, (1.56 x 10 to 7.30 x 10 s ), as well as escape rate constants (7.S5 x 10 to 3.*»4 x 10 s ) and diffusion coefficients (3.39 x 10" to 0-.88 x 10" cm'Vs) for these species in the fuel-to-sheath gap. Analyses al.'.o enable identification of the various rate- controlling processes operative in each element. For the noble gas and iodine species, the rate-determining process in the multi-slit element is "Booth diffusion"; however, for the hydrided element an additional delay results from diffusional transport in the fuel-to-sheath gap. Furthermore, the iodine species exhibit an additional holdup in the drilled element because of significant trapping on the fuel and/or sheath surfaces. Using experimental release data and applying the theoretical results of this wori<, a systematic procedure is proposed to characterize fuel failures in commercial power reactors (i.e., the number of fuel failures and average leak size). ACKNOWLEDGEMENTS This thesis ,va-, carried <mt mder the supervision of Prof. C.R. Phillips ('J of T) and M.7.F. Motley (GkNL) whose advice and support are gratefully acknowledged. Tiie author would like to especially thank Professors D.G. Andrews, R.A. Bonalumi and R.E. Jervis for their continuing nelp and guidance. The drilled defect experiment (FOO-&SI) was carried out by R.D. MacDonald at CRNL. Although the hydrided defect experiment (PFO-102-2) and the muiti-siit defect experiment (FFO-103) were carried out under my direct control and supervision, I would like to acknowledge the following assistance at CRNL. I wish to thank D.G. Hardy and D.R. McCracken for their helpful suggestions and assistance, and Y.H. Kang of the Korea Advanced Energy Research Institute for assistance with the preliminary data analysis. Thanks are also due to E.P. Penswick, D.D. Semenkik, D.F. Shields and various members of the Reactor Loops Branch l^r iheir conscientious data acquisitioning and processing, and to 3. Blair, M. Milgrom and R.L. da Silva for their invaluable help with the computer codes (see Appendix A). ! would aJso (ike to express my gratitude to R.l. Chenier, A.f:. Hut ton, !. Lu-I< and R.E. Moeller for their comprehensive rnetallographic examination, .I.E. Winegar for the X-ray diffra<;tion analysis and D.A. Leach for the neutron radiography examination. Special thanks are due to L. Blimkie and G.R. Phillips for their superb tracing work. Finally, ! would like to express my sincere thanks to my wife Patricia for her continuing help and support throughout this endeavour. Part of this study was funded by the Ontario Hydro - Atomic Energy of Canada Limited 3oint Programme. CONTENTS Pat-e 1. Introduction ' 2. Historical Review J 2.1 Previous Defect Models 5 2.2 Experimental Background 7 3. Experimental Detaiis '-> 3.1 Loop Monitoring Facililty 9 3.2 Experiment Description LI 3.2.1 Fuel Element Dpsign and Previous Irradiation History 11 3.2.2 Loop Operating Conditions li 3.2.3 Irradiation History J1 3.3 Post-Irradiation Examination 16 3.4 Discussion of Sheath and Fuel Degradation 22 4. Steady-State Fission Product Release 24 4.1 Results and Analyses 24 4.2 Discussion of Noble Gas and Iodine Behaviour 31 4.2.1 Chemical Nature of Iodine .31 4.2.2 Sweep Gas Experiments 33 4.2.3 De'L'Ct Experiments '34 4.3 Semi-rlmpirical Fission Product Release Model 37 4.3.1 Diifusion in the UO_ Fuel 37 4.3.2 Transport in die Fuel-to-Sheath Gap 42 4.3.3 Diffusion in i.!i<: Riel-to-Sheath Gap 48 4.4 Evaluation of Model Parameters 54 4.5 Discussion of Empirical Results 58 4.6 Comparison of Model with Previous Models 64 5. Transient Fission Product Release 67 5.1 Results 67 5.2 Discussion 70 6. Power Reactor Applications 72 7. Conclusions and Summary of Results 79 8. Recommendations 82 9. Nomenclature 83 10. References 85 IV Appendix A Computer Code Description 90 A.I The GRAAS Code 91 A.2 The SUMRT Code 9? A.3 The LATRIII' Code <,', ,\A The FISSPROD Code g7 A.5 The MLSQQ Code o8 Appendix ii The X-2 Loop Mode! 100 Appendix C Ftooth Diflusionj] Release Corrected for Precursor Diffusion 102 Appendix D Diffusion in the FueJ-to-Sheath Gap JD/J Appendix E Rare Gas Diffusion Coefficient from Kinetic Theory 1()a Appendix F Fuel Element Temperature Calculations -|11 Appendix G Thermodyjuriiir Calculations I [7 Appendix H Selected ENDF/1VV Decay Data no Appendix I Loop Design and Measurement 122 1.1 Description of the Loop 122 1.2 Spectrometer Design 123 1.3 Discussion of Error 124 LIST OF TABLES AND FIGURES Tables 3.1 Pre-Irradiation Fuel Element Design 1.! 3.2 Comparison of the X-2 Loop Coolant Conditions with the Specified Bruce NGS Primary Meat Transport System Conditions ] ;< 3.3 Average Element Power Ratings, Full Power Days and Ournups 13 3.4 Oxygen-to-Metai Ratios (O/U) for Elements A7E and A3N IS 3.5 Zirconium Oxide Thicknesses on the Sheath Surfaces -'1 4.1 Release Rate (R ), Birth Rate (i\) and Fractional Release (F) Values During Steady Reactor Power for Experiments FDO-6S1, FFO-102-2and FFO-103 ^6 4.2 Fractional Release Dependency on the Decay Constant of the Form a.\° 30 4.3 Comparison of the Slope on a Fractional Release Versus Decay Constant Plot for Various Defect Experiments 35 4.4 Correction Fa-.tor (H) ior Pr-.-rursor Diffusion '•, I 4.5 Evaluation of Empirical Par.n-'icters for FOO-6S1, FFO-102-2 and FFO-103 35 4.6 Experimental Gap Diffusiv', :ies for FOO-6S1 and FFO-102-2 56 4.7 Comparison of Measured and Calculated Fractional Releases 5 7 4.8 Comparison of Theoretical and Experimental Noble Gas Diffusivitics for FDO-W! and FFO-102-2 61. 4.9 Comparison of Calculated Fractional Releases (F) for Various Defect Models 64 6.1 Isotopic Release Rate Ratios for Experiments FDO-6S1, FFO-102-2 and FFO-103 and for a Recoil System 75 6.2 Release Rate Dependencies on the Decay Constant for Particular Reactor Core Conditions 76 6.3 1-131 Release Rates for Experiments FDO-681, FFO-102-2 and FFO-103 77 E.I Calculation of Rare Gas Diffusion Coefficients from Kinetic Theory 109 F.I Symbols Used in Temperature Calculation 113 G.I Calculation of Free Energies at 623 K (350 °C) for the Ionic Species 118 1.1 The X-2 Loop Design Parameters 122 1.2 Detector/Collimator Combination for Experiments FDO-681 FFO-102-2 and FFO-103 124 3.1 The X-2 Defect Loop !(l 3.2 Experiment FFO-102--22 Reactor Power History and Fission Product Concentratiomn and Release Rate Historvy for !<r-85m 1^ 3.3 Visual and Metal lographic Examinations of Element A7E 17 3.4 Post-Irradiation Visual and Mo tal lographic Examinations of Element A3N from FFO-103 !'-> 3.5 Grain Boundary Separation Indicating Oxidation of the Futl Surface at the Main Defect Site for Element A7E -° 4.1 Fractional Release Versus Decay Constant Plots for FDO-6SI, FFO-102-2 and FFO-103 '"> 4.2 iodine Species Distribution Diagrams ^ 4.3 Rare Gas Diffusion Coefficient as a Function of Temperature and Stoichiometry '$ 4.4 Correlation of Empirical Diffusion Coefficients Versus Linear Heat Rating for Intact and Defective German BWR and PWR Fuel Rods 59 5.1 Fission Product Activity Concentrations During a Reactor Trip for FFO-102-2 69 5.2 Iodine, Xenon, Krypton and Cesium Concentrations During a Scheduled Reactor Shutdown for FFO-102-2 70 D.I Coordinate System for a Defect in a Cylindrical Fuel Element 104 E.I Generalized Chart for Self-Diffusivities of Dense Gases 110 !. INTRODUCTION For over a decade fuel performance in the Canadian CANDU reactor system has been excellent. Remote instances of unsatisfactory performance were attributed to inadequate quality control procedures during fuel fabrication. Fuel elements containing manufacturing, porous end plug and faulty end weld defects (such that a small leak path is present between the primary coolant and the internal element atmosphere) have exhibited secondary sheath deterioration following irradiation.
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