Generalized Dancoff Factor in Complex Lattice Arrangement

Protected Pu production (P³) is a recently developed concept to enhance the proliferation resistance properties of Pu by increasing the fraction of even-mass-number Pu isotopes thus leading to the denaturing of Pu. The present paper deals with the possibility of denaturing Pu by consideration of Pu and minor actinides (MAs) multicycling in medium-size fast breeder reactors (FBRs). It was found that multicycling of Pu and MAs in FBRs enhances the proliferation resistance of Pu by increasing the fraction of evenmass-number Pu isotopes. The proliferation resistance of Pu in P³ recycling options satisfied criteria based on Kessler's proposal, i.e., 6% ²³⁸Pu content, and Pellaud's proposal, i.e., 30% ²⁴⁰Pu content. The normalized attractiveness of Pu in P³ recycling options satisfied both criteria in the core as well as in the blanket with the combination of even-mass-number Pu isotopes. At the initial introduction, two P³ FBRs (714 MWth class) with the addition of 5 wt% MA into the blanket can be operated using the annual discharged spent fuel in Japan.     

Decrease of Cesium Release from Irradiated UO2 Fuel in Helium Atmosphere under Elevated Pressure of 1.0 MPa at Temperature up to 2,773K

In the case of severe accidents, the radionuclides release from fuel could mostly occur at high temperature under elevated pressure. The effect of temperature on the release has been clarified in many previous studies while the pressure influence has been scarcely investigated so far due to difficulty in the experimental operation. To investigate the effect of pressure on the release, two tests under the same conditions except for the system pressure were performed in the VEGA program at JAERI by heating up the irradiated UO₂fuels up to 2,773 K in inert helium. The test results uniquely showed that the release rate of cesium for the temperatures below 2,773 K at 1.0 MPa could be suppressed by about 30% compared with that at 0.1 MPa. This article describes the outlines of the two tests and the observed effects of system pressure on cesium release as well as the results of various post-irradiation examinations. Moreover, the mechanisms and models that explain the pressure effect are discussed.     

Measurements of Doppler Effect of Coated Particle Fuel Rod in SHE-14 Core Using Sample Heating Device

Reactivity decrease due to temperature rise of a single fuel rod sample was measured in the SHE-14 core using a sample heating device with purpose to verify the calculation accuracy of the Doppler effect for resonance neutron absorption in a Very High Temperature Reactor. The fuel rod sample was a stuck of coated particle fuel compacts containing 4% enriched UO₂ kernels, and it was heated up to about 700°C in a sample heating tube which was placed along the axis of the core.

The difference of reactivity decrease between the two same size samples of fuel rod and graphite rod due to temperature rise can be interpreted as the increased resonance neutron capture of ²³⁸U, i.e. Doppler effect. The SRAC code system was applied to the Doppler effect calculation where the collision probability method was used in the cell calculation and the one-dimensional, multi-group diffusion approximations were adopted in the core calculation. The results gave a ratio of the calculated to the measured Doppler effect of 0.93.     

Fabrication of Sintered Annular Fuel Pellet for HANARO Irradiation Test

The fabrication method of an annular pellet with highly precise diametric tolerances, same dimensions, and various sintered densities has been investigated. To examine the in-pile densification and swelling of the annular pellet, 5 different types of annular pellet were prepared for a HANARO irradiation test. In order to obtain annular fuel pellets with the same dimensions and various sintered densities, we control the green density of an annular compact, the sintering temperatures, and the sintering time. For a diametric tolerance control, we have introduced a new compaction process that combines the usual double-acting pressing and cold isostatic pressing. Annular fuel pellets with the same dimensions and various sintered densities were fabricated successfully, and all the pellets satisfied the pellet specification of the HANARO irradiation test. Sintered annular pellets show an excellent inner diametric tolerance of less than ±12 μm without an inner surface grinding.     

Lattice Parameter Measurements on Cluster-Type Fuel for Advanced Thermal Reactor

Abstract

The behavior of neutrons in a highly heterogeneous unit cell consisting of D₂O moderator, H₂O coolant and a 28-pin fuel cluster contained in a pressure tube has been studied through lattice parameter measurements covering three different ²³⁵U enrichments, four coolant void fractions and two lattice pitches. A single-region core configuration was adopted, with which measurements were made to determine—in relation to coolant void fraction—the critical D₂O level, as well as various lattice parameters

A strong dependence on coolant void was observed for the critical level and the lattice parameters, in the case of the smaller 22.5 cm pitch lattice, due to the positive effect on core reactivity exerted by the slowing-down faculty of H₂O in the epithermal energy region. With the larger 25.0 cm pitch, however, no meaningful dependence on void fraction was shown by any of the measured values, and this was ascribed to a compensating negative effect due to enhanced thermal neutron self-shielding in the fuel region produced by the H₂O coolant.

The results of cell calculations obtained by means of the METHUSELAH-II code showed generally good agreement with experimentally determined data, for both critical D₂O levels and lattice parameters, in the case of coolant-filled lattices (0, 30 and 70% void fractions). For lattices devoid of coolant (100% void fraction), however, discrepancies in lattice parameters—particularly in p ²⁸—produced corresponding deviations in core reactivity amounting to 1% in excess of those incurred with other void fractions.     

Effects of Void Migration Process on Fission-Gas Release

The release rate of fission-gas from U0₂ was continuously measured during irradiation in the Hitachi Training Reactor. The U0₂ specimen was heated electrically in in-core assemblies by tungsten heaters, either arranged axially transversing the specimen (producing radial temperature gradient) or cylindrically outside the specimen (uniform heating). In a case of the axially heated annular U0₂ pellets with radial temperature gradient, the rate of fission-gas release increased with time under constant temperature and flux. This experimental result was explained by considering as fission gas release mechanism a combination of diffusion and pore migration processes. On the other hand, when U0₂ pellets were heated isothermally up to 2,000°C, the fission-gas release rate could be explained in terms of diffusion process alone. No effect of equiaxial grain growth was observed on the rate of fission-gas release.     

A New Aspect of Grain Growth of Sol-Gel Urania

Grain growth in sol-gel urania compacts was determined as a function of temperature by observing with scanning electron microscope the fractured surface of compacts heated up to predetermined temperatures at a constant rate of heating. Rapid grain growth, unexplainable by currently accepted square or cubic growth kinetics, was observed to occur in the sintered compact when the density reached 70 to 75%T.D., which was attained at about 1,000°C in the case of heating rate of 5°C/min. It was considered that this rapid grain growth proceeded by plastic flow of which driving force was the surface tension of the grains, when a large number of primary grains were closely packed against each other, i.e., arranged in quite high coordination number. A rough estimation showed that the rate of coalescence of grains was dependent on the temperature and the grain size or the surface tension of the material, and that the multi-grain-coalescence model, which was previously reported for the case of sintering of BeO, could also be applied to sol-gel urania.     

Thermal Diffusion Factor for Diatomic Gas Mixture in Multicomponent System

Thermal diffusion factor for diatomic gas mixture was expressed in a practical form applicable to multi- component mixture. Thermal diffusion factor for diatomic gas mixtures was obtained directly from thermal diffusion coefficient in multicomponent mixture by using the relation between thermal diffusion coefficient and thermal diffusion factor, and was shown to be the same as that derived by Mason et al. through a “generalized Stefan-Maxwell equation”. The present expression gives a theoretical proof of the previously reported rough estimate of thermal diffusion factor for hydrogen binary mixtures. Comparing with the present expression for H2-HT binary gas mixture, we have found that the previously reported expression for thermal diffusion factor overestimates in a temperature range from 50 to 1,200 K, but the difference between the previous and the present expressions is small. The present expression is essential to the separative analyses of thermal diffusion column for diatomic gas mixture in multicomponent system, and is consistent with the rough estimate for binary mixture of diatomic gases.     

亚化学计量UO2-x燃料芯块的制备机理

亚化学计量UO_2-x芯块是一种设计新颖的特殊核反应堆用核燃料,很难采用传统压水堆超化学计量UO_2+x+U芯块工艺进行制造。本工作采用UO_2+x+U混合粉末为原料制备了UO_2-x芯块,研究了铀粉表面包覆处理方法、铀粉含量、成型压力、烧结气氛等工艺参数对芯块O/U比、烧结密度和微观结构的影响,探讨了UO_2-x环形芯块的亚化学计量形成机理。研究表明,当铀粉加入量（质量分数）分别为0、3%、6%时,芯块O/U比分别为2.010、1.991、1.982,平均晶粒尺寸分别为10、15、20μm;当铀粉加入量为50%时,O/U比为1.943,样品发生熔化。亚化学计量UO_2-x芯块必须在干燥惰性气氛中密封保存。     

Effects of Temperature Cycling and Heating Rate on Fission Gas Release of BWR Fuels

The effects of temperature cycling and heating rate on the release behavior of ⁸⁵Kr have been studied for U0₂ pellets irradiated in a commercial BWR during 3 and 4 cycles (burn-up: 23 and 28GWd/t), by using a post irradiation annealing technique. In addition, characteristics of intergranular bubbles in base-irradiated and annealed specimens (burn-up: 6~28GWd/t) have been examined by SEM fractography.

No significant difference in the release of ⁸⁵Kr was observed between the cyclic heating from 700 to 1,400°C and isothermal heating at 1,400°C. The maximum release rate of ⁸⁵Kr during heating up to 1,800°C became lower with decreasing heating rate in the range of 0.03–10°C/s, while its cumulative fractional releases were about 20~30%, almost independent of heating rate. The fractional coverage of the grain face area occupied by intergranular bubbles saturated around 40~50 for the specimens annealed at 1,600-1,800°C, independent of specimen burn-up and heating conditions (temperature, heating rate and duration). A relationship between intergranular bubble concentration N_g per unit area of grain face and average bubble diameter d_g was expressed as N_g∝d_{g 2.1}     

Hydrogen Concentration Dependence of Tritium Tracer Diffusion Coefficient in Alpha Phase of Niobium

In order to examine influences of coexistent hydrogen isotopes on diffusion behavior of tritium in niobium, tracer diffusion coefficients D_t of tritium in alpha phase of hydrogenated and deuterized niobium (α-NbH_xT_y and α-NbD_xT_y (x<0,8,y<<x)) have been measured at 473 K, 493 K and 553 K. The data on D_t show typical hydrogen concentration dependence: D_t of tritium for both α-NbH_xT_y, and α-NbD_xT_y, decreases with hydrogen concentration under all experimental conditions. The obtained concentration dependence of D_t of tritium differs from that of D_t of protium in α-NbH_x or of deuterium in α-NbH_xT_y. On the other hand, no appreciable differences in the concentration dependence of D_t of tritium between α-NbH_xT_y, and α-NbD_xT_y, are observed: there are no definite isotope effects due to the coexisting hydrogen isotopes. This result suggests that D_t of tritium for a tritiated niobium (α-NbT_x) is not very different from that for α-NbH_xT_y and α-NbD_xT_y. The chemical diffusion coefficient D* of tritium is also evaluated on the basis of the obtained D_t of tritium and of a literature value of a thermodynamic factor F for Nb-H and Nb-D systems.     

Heat Capacity of Thorium Nitrides from 450 to 850K

The throium nitrides (ThN and Th₃N₄) were prepared by solid-gas reaction with thorium hydride and nitrogen.

The heat capacity of these samples were measured from 450 to 850K by a high-temperature double- adiabatic calorimeter and were determined as functions of temperature as follows:

Th₃N₄: C _p=41.30–8.47×10^?3 T-6.37×10⁵ T ^?2

ThN: C _p=12.50–2.66×10^?3 T-2.44×10⁵ T ^?2