Accident analysis computer code for nuclear reactors

Originally presented as the first author's thesis, (Nucl. E.)--in the M.I.T. Dept. of Nuclear Engineering, 1977Includes bibliographical references (pages 122-123

Design of central irradiation facilities for the MITR-II research reactor

"September 1976."Also issued as a Ph. D. thesis by the first author and supervised by the second author, MIT Dept. of Nuclear Engineering, 1977Includes bibliographical references (pages 94-95)Design analysis studies have been made for various in-core irradiation facility designs which are presently used, or proposed for future use in the MITR-II. The information obtained includes reactivity effects, core flux and power distributions, and estimates of the safety limits and limiting conditions for operation. A finite-difference, diffusion theory computer code was employed in two and three dimensions, and with three and fifteen group energy schemes. The facilities investigated include the single-element molybdenum sample holder, a proposed double-element irradiation facility and a proposed central irradiation facility design encompassing most of the area of the three central core positions. In addition, a comparison of the effects of various absorber materials has been made for a core configuration which includes three solid dummies. Flux levels in the molybdenum holder facility and in the beam ports were calculated for both three and five dummy cores. Flooding the sample tube in these cases was found to increase the safety and operating limits, but not to unacceptable levels for an 8 inch blade height. For the five dummy case, the operating limit in the C-ring was predicted to reach its maximum allowed value at a blade bank height of 13.6 inches. The reactivity effect of flooding was calculated to be 0.19%AK for the five dummy case, in direct agreement with the measured value. Flooging the large sample channel in the double element facility was found to increase the reactivity by 1.5 6%AK ff and also to cause an unacceptable power-peaking. The proposed central irradiation facility is a thermal flux-trap which could produce thermal flux values of up to 2.0 x 1014 n/cm 2 sec. Computer estimates show that flooding this facility's central sample tube would increase this value to 2.5 x 1014 n/cm2 sec, without resulting in an unacceptable power peak

An evaluation of tight-pitch PWR cores

Originally presented as the author's thesis, Ph.D. in the M.I.T. Dept. of Nuclear Engineering, 1979.The impact of tight pitch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system-U-235/U02: Pu/Th02: U-233/ThO2--and the conventional recycle-mode uranium system- U-235/U02: Pu/UO . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) o the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices)< F/M < 4.0 are limited by the scarcity of experiments with F/M > l.0,the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments. It was found that by increasing F/M to "3 the uranium ore usage for the uranium system can be decreased by as much as 60% compared to the same system with conventional recycle (at F/M 0.5). Equivalent savings for the thorium system of the type examined here are much smaller (10%) because of the poor performance of the intermediate Pu/ThO2 core--which is not substantially improved by increasing F/M. Although fuel cycle costs (calculated at the indifference value of bred fissile species) are rather insensitive to the characteristics of the tight pitch cores, system energy production costs do not favor the low discharge burnups which might other- wise allow even greater ore savings (80%). Temperature and void coefficients of reactivity for the tight pitch cores were calculated to be negative. Means for implementing tight lattice use were investigated, such as the use of stainless steel clad in place of zircaloy; and alternatives achieving the same objective were briefly examined, such as the use of D20/H20 mixtures as coolant. Major items identified requiring further work are system redesign to accommodate higher core pressure drop, and transient and accident thermal-hydraulics.DOE Contract no. EN-77-S O2-4570

Heterogeneous effects in fast breeder reactors

"January, 1973."Also issued as a Ph. D. thesis written by the first author and supervised by the second and third author, MIT, Dept. of Nuclear Engineering, 1973Includes bibliographical references (pages 259-266)Heterogeneous effects in fast breeder reactors are examined through development of simple but accurate models for the calculation of a posteriori corrections to a volume-averaged homogeneous representation. Three distinct heterogeneous effects are considered: spatial coarse-group flux distribution within the unit cell, anisotropic diffusion, and resonance self-shielding. An escape/transmission probability theory is developed which yields region-averaged fluxes, used to flux-weight cross sections. Fluxes calculated by the model are in substantial agreement with S 8 discrete ordinate calculations. The method of Benoist, as applied to tight lattices, is adopted to generate anisotropic diffusion coefficients in pin geometry. The resulting perturbations from a volume-averaged homogeneous representation are interpreted in terms of reactivities calculated from first order perturbation theory and an equivalent "total differential of k" method.Resonance self-shielding is treated by the f-factor approach, based on correlations developed to give the self-shielding factors as functions of one-group constants. Various reference designs are analyzed for heterogeneous effects. For a demonstration LMFBR design, the whole-core sodium void reactivity change is calculated to be 90e less positive for the heterogeneous core representation compared to a homogeneous core, due primarily to the effects of anisotropic diffusion. Parametric studies show at least a doubling of this negative reactivity contribution is attainable for judicious choices of enrichment, lattice pitch and lattice geometry (particularly the open hexagonal lattice). The fuel dispersal accident is analyzed and a positive reactivity contribution due to heterogeneity is identified. The results of intra-rod U-238 activation measurements in the Blanket Test Facility are analyzed and compared to calculations.This activation depression is found to be of the order of 10% (surfaceto- average) for a typical LMFBR blanket rod and is ascribed to the effect of heterogeneous resonance self-shielding of U-238. Heterogeneous effects on the breeding ratio are studied with the conclusions that accounting for resonance self-shielding reduces the total breeding ratio by over 10%, but heterogeneous effects are not important for breeding ratio calculations.U.S. Atomic Energy Commission contract AT (11-1) - 225

Analysis of strategies for improving uranium utilization in pressurized water reactors

Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal objective has been the evaluation of suggested improvements on a self-consistent basis, allowing for concurrent changes in dependent variables such as core leakage and batch power histories, which might otherwise obscure the sometimes subtle effects of interest. Two levels of evaluation have been devised: a simple but accurate analytic model based on the observed linear variations in assembly reactivity as a function of burnup; and a numerical approach, embodied in a computer program, which relaxes this assumption and combines it with empirical prescriptions for assembly (or batch) power as a function of reactivity, and core leakage as a function of peripheral assembly power. State-of-the-art physics methods, such as PDQ-7, were used to verify and supplement these techniques.These methods have been applied to evaluate several suggested improvements: (1) axial blankets of low-enriched or depleted uranium, and of beryllium metal, (2) radial natural uranium blankets, (3) lowleakage radial fuel management, (4) high burnup fuels, (5) optimized H/U atom ratio, (6) annular fuel, and (7) mechanical spectral shift (i.e. variable fuel-to-moderator ratio) concepts such as those involving pin pulling and bundle reconstitution.The potential savings in uranium requirements compared to current practice were found to be as follows: (1) O0-3%, (2) negative, (3) 2-3%; possibly 5%, (4) "15%, (5) 0-2.5%, (6) no inherent advantage, (7) 10%. Total savings should not be assumed to be additive; and thermal/hydraulic or mechanical design restrictions may preclude full realization of some of the potential improvements

An accident probability analysis and design evaluation of the gas-cooled fast breeder reactor demonstration plant

Originally presented as the first author's thesis (Ph. D.), M.I.T. Dept. of Nuclear Engineering, 1976Includes bibliographical references (pages 510-515

Reactor physics calculation of BWR fuel bundles containing gadolinia

"January 1977.""YAEC-1126."Includes bibliographical references (pages 142-144)A technique for the calculation of the neutronic behavior of BWR fuel bundles has been developed and applied to a Vermont Yankee fuel bundle. The technique is based on a diffusion theory treatment of the bundle, with parameters for gadolinia bearing pins generated by transport theory, and converted to effective diffusion- theory values by means of blackness theory. The method has been used to examine the dependence of various bundle average parameters on control rod insertion history

Analysis of strategies for improving uranium utilization in pressurized water reactors

Includes bibliographical references (pages 238-241)Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal objective has been the evaluation of suggested improvements on a self-consistent basis, allowing for concurrent changes in dependent variables such as core leakage and batch power histories, which might otherwise obscure the sometimes subtle effects of interest. Two levels of evaluation have been devised: a simple but accurate analytic model based on the observed linear variations in assembly reactivity as a function of burnup; and a numerical approach, embodied in a computer program, which relaxes this assumption and combines it with empirical prescriptions for assembly (or batch) power as a function of reactivity, and core leakage as a function of peripheral assembly power. State-of-the-art physics methods, such as PDQ-7, were used ! to verify and supplement these techniques.These methods have been applied to evaluate several suggested improvements: (1) axial blankets of low-enriched or depleted uranium, and of beryllium metal, (2) radial natural uranium blankets, (3) low-leakage radial fuel management, (4) high burnup fuels, (5) optimized H/U atom ratio, (6) annular fuel, and (7) mechanical spectral shift (i.e. variable fuel-to-moderator ratio) concepts such as those involving pin pulling and bundle reconstitution.The potential savings in uranium requirements compared to current practice were found to be as follows: (1) O0-3%, (2) negative, (3) 2-3%; possibly 5%, (4) "15%, (5) 0-2.5%, (6) no inherent advantage, (7) 10%. Total savings should not be assumed to be additive; and thermal/hydraulic or mechanical design restrictions may preclude full realization of some of the potential improvements

Design and fuel management of PWR cores to optimize the once-through fuel cycle

DOE Contract no. EN-77-S-02-4570Originally presented as the first author's thesis, (Sc. D.)--in the M.I.T. Dept. of Nuclear Engineering, 1978Includes bibliographical references (pages 238-241

The reactor physics of the Massachusetts Institute of Technology reactor redesign

"August, 1970."Also written as a Ph. D. thesis by the first author and supervised by the second and third author, MIT, Dept. of Nuclear Engineering, 1970Includes bibliographical references (pages 284-289)An H20 cooled compact MITR-II core, reflected by D20 has been designed for the MITR to increase the reflector thermal neutron flux at tips of beam ports by a factor of 3 or better, without changing the operating power level of the reactor. The diffusion approximation to the neutron transport equation has been used. A three neutron energy group scheme, that retains essential spatial effects, used in the studies has yielded satisfactory agreement with measured data. The factors which affect the intensity as well as the quality of the reflector thermal neutron flux have been studied. These studies show that the permanent features of the MITR limit the maximum power densities in the MITR-II core to factors between 4.5 and 12 below the corresponding values in reactors employing a similar core concept Nevertheless, the predicted unperturbed reflector thermal neutron flux of 1.lXlO14 n/cm 2-sec in MITR-II yields a reflector flux per unit power that is competitive with the corresponding values available in reactors of its type and a factor of 5.0 higher than that in MITR-I