Observation of Porosity Reduction in a Densification-Prone Test Fuel Rod: Data and Analysis

Instrumented fuel assembly (IFA)-431 was irradiated in the Halden Boiling Water Reactor (HBWR) for the purpose of extending the steady-state data base. Rod 6 of this assembly began irradiation with UO{sub 2} fuel of 92% theoretical density (TD) that was unstable with respect to in-reactor densification. Thermal resintering tests resulted in a final density of 95.3% TD while post-irradiation examination (PIE) indicated a final density of 96.5% TD. Observed microstructural changes were consistent with published densification studies; there was a marked depletion of submicrometer diameter pores and total pore volume. However, grain size increased only slightly, indicating that internal pellet temperatures did not reach the 1875K applied in resintering tests. Oensification was observed to increase the temperatures in rod 6, but temperatures did not become as high as for a sibling rod that simulated instantaneous densification. Temperatures calculated with U.S. Nuclear Regulatory Commission (NRC) fuel performance computer codes were generally higher than observed temperatures

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

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

Originally presented as the first author's thesis, (Sc.D.) in the M.I.T. Dept. of Nuclear Engineering, 1978.The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current light water reactors, with a specific focus on pressurized water reactors. The types of changes which have been examined are: (1) re-optimization of fuel pin diameter and lattice pitch, (2) Axial power shaping by enrichment gradation in fresh fuel, (3) Use of 6-batch cores with semi-annual refueling, (4) Use of 6-batch cores with annual refueling, hence greater extended (.doubled) burnup, (5) Use of radial reflector assemblies, (6) Use of internally heterogeneous cores (simple seed/blanket configurations), (7) Use of power/temperature coastdown at the end of life to extend burnup, (8) Use of metal or diluted oxide fuel, (9) Use of thorium, and (10) Use of isotopically separated low a cladding material. a State-of-the-art LWR computational methods, LEOPARD/PDQ-7/FLARE-G, were used to investigate these modifications. The most effective way found to improve uranium ore utilization is to increase the discharge burnup. Ore savings on the order of 20% can be realized if greatly extended burnup (- double that of current practice) is combined with an increase in the number of batches in the core from 3 to 6. The major conclusion of this study is that cumulative reductions in ore usage of on the order of 30% are fore- seeable relative to a current PWR operating on the once-through fuel cycle, which is comparable to that expected for the same cores operated in the recycle mode.DOE Contract no. EN-77-S-02-4570

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 evaluation of the fast-mixed spectrum reactor

"February 1980."Also issued as an M.S. thesis written by the first author and supervised by the second and third authors, MIT Dept. of Nuclear Engineering, 1980Includes bibliographical references (pages 145-147)An independent evaluation of the neutronic characteristics of a gas-cooled fast-mixed spectrum reactor (FMSR) core design has been performed. A benchmark core configuration for an early FMSR design was provided by Brookhaven National Laboratory, the originators of the concept. The results of the evaluation were compared with those of BNL. Points of comparison included system reactivity and breeding ratio, and region-wise power densities and isotopic compositions as a function of burnup. The results are in sufficiently good agreement to conclude that the neutronic feasibility of the FMSR concept has been independently validated. Significant differences, primarily in higher plutonium isotope concentrations, occur only in regions of low neutronic importance, and plausible reasons for the differences are advanced based on sensitivity studies and comparison of spectral indices. While both M.I.T. and BNL calculations tend to predict that the benchmark design is slightly subcritical, at the beginning of equilibrium cycle, the margin to k = 1.0 is close enough (Ak < 0.03) that the situation can be remedied. Establishment of a consensus fission product cross section set was identified as an objective of merit, since non-negligible differences were found in results computed using various extant sets (BNL, LIB-IV, Japanese). Non-fission heating by gamma and neutron interactions was evaluated for the reference core design using a coupled neutron/gamma cross section set and SN calculations. In the unfueled regions of the core, moderator elements in particular, the non-fission heating rate was found to be significant (averaging about 6 kw/liter), but posed no obvious problems. In fueled regions the common assumption of local deposition of all energy at the point of fission was verified to be a good approximation for most engineering purposes.Engineering and Advanced Reactor Safety Division of the U.S. Department of Energy at Brookhaven National Laboratory contract 472241-

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

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

Optimization of the axial power shape in pressurized water reactors

Statement of responsibility on title-page reads: M.A. Malik, A. Kamal, M.J. Driscoll, and D.D. Lanning"November 1981."Originally presented as the first author's M.S. thesis, M.I.T. Dept. of Nuclear Engineering, 1981Includes bibliographical references (pages 112-114)Analytical and numerical methods have been applied to find the optimum axial power profile in a PWR with respect to uranium utilization. The preferred shape was found to have a large central region of uniform power density, with a roughly cosinusoidal.profile near the ends of the assembly. Reactivity and fissile enrichment distributions which yield the optimum profile were determined, and a 3-region design was developed which gives essentially the same power profile as the continuously varying optimum composition. State of the art computational methods, LEOPARD and PDQ-7, were used to evaluate the beginning-of-life and burnup history behavior of a series of three-zone assembly designs, all of which had a large central zone followed by a shorter region of higher enrichment, and with a still thinner blanket of depleted uranium fuel pellets at the outer periphery. It was found that if annular fuel pellets were used in the higher enrichment zone, a design ! was created which not only had the best uranium savings (2.8% more energy from the same amount of natural. uranium, compared to a conventional, uniform, unblanketed design), but also had a power shape with a lower peak-to-average power ratio (by 16.5%) than the reference case, and which held its power shape very nearly constant over life. This contrasted with the designs without part length annular fuel, which tended to burn into an end-peaked power distribution, and with blanket-only designs, which had a poorer peak-to-average power ratio than the reference udblanketed case.DOE contract no. DE-AC02-79ET340

Use of neutron absorbers for the experimental determination of lattice parameters in subcritical assemblies

Statement of responsibility on title-page reads: J. Harrington, D. D. Lanning, I. Kaplan, T. J. Thompson"February 1966."AEC Research and Development ReportMIT-2344-6Includes bibliographical references (leaves 188-192)U.S. Atomic Energy Commission contract AT(30-1)234