Takenori Suzaki

Non-Destructive Gamma-Ray Spectrometry on Spent Fuels of a Boiling Water Reactor

The spent fuels from the JPDR-I reactor were measured by means of a γ-scanning facility installed in the fuel storage pool. The spatial distributions of the fission products (134Cs and 137Cs) were measured and analyzed in reference to the effects of control rod pattern. The ratios holding between the products of neutron capture and of direct fission (134Cs/137Cs and 154Eu/137Cs) were also examined for its relevance to non-destructive burnup determination. The activity ratios of the fission products can be expressed by a linear function of burnup, provided that corrections are made to account for differences in irradiation history and for spatial variations in the neutron spectrum.

Measurements of the Modified Conversion Ratio by Gamma-Ray Spectrometry of Fuel Rods for Water-Moderated UO2 Cores

The modified conversion ratio is defined as the ratio of 238U captures to total fission. Gamma-ray spectrometry of irradiated fuel rods has been introduced to measure this quantity in two types of ...

Subcriticality Determination of Low-Enriched UO2 Lattices in Water by Exponential Experiment.

To determine the static k (effective neutron multiplication factor) ranging from the critical to an extremely subcritical states, the exponential experiments were performed using various sizes of light-water moderated and reflected low-enriched UO2 lattice cores. For comparison, the pulsed neutron source experiments were also carried out. In the manner of the Gozanis bracketing method applied to the pulsed source experiment, a formula to obtain k from the measured spatial-decay constant was derived on the basis of diffusion theory. Parameters in the formulas needed to obtain k from the respective experiments were evaluated by 4-group neutron diffusion calculations. The results of the exponential experiments agreed well with those of the pulsed source experiments, the 4-group diffusion calculations and the 137-group Monte Carlo calculations. Therefore, the present data-processing method developed for the exponential experiment was demonstrated to be valid. Besides, through the examination on the parameter...

Determination of the modified conversion ratio of light-water-moderated uranium-plutonium mixed-oxide-fuel lattice

The modified conversion ratio (MCR) (the ratio of the {sup 238}U capture rate to the total fission rate) in a light-water-moderated uranium-plutonium mixed-oxide (MOX-) fuel lattice was measured for four types of lattices with different plutonium enrichment. In the current method, the relative reaction rates of {sup 238}U capture and total fission were obtained from nondestructive gamma-ray spectrometry of {sup 239}Np and fission products, respectively, which accumulated in the fuel rod irradiated at the Tank-Type Critical Assembly. The measured results of the fission rates derived from two different fission products agreed well with each other, and the measured MCRs showed good agreement with the results of the Monte Carlo calculation with the whole-core model. Therefore, the current nondestructive method is applicable to the MCR measurement of MOX fuel.

Study on the Criticality Safety Evaluation Method for Burnup Credit in JAERI

In relation to burnup credit, three tasks have been carried out at the Japan Atomic Energy Research Institute (JAERI) for establishing the evaluation method of criticality safety for a spent-fuel system, such as storage ponds and transports casks. The first task is to prepare a brenchmark database of criticality experiments and nuclide compositions of spent fuels. The database of nuclide composition is formed by data measured at JAERI and data collected from the literature. For the database of criticality experiments, the effective multiplication factor of a spent-fuel assembly has been measured at JAERI. The next task is to develop computer codes. The burnup and criticality codes have been developed and validated by analyzing a large number of benchmarks stored in the aferomentioned database. The last task needed to establish the methodology in order to confirm the subcriticality of a spent-fuel system applying burnup credit described. A reference fuel assembly is introduced so that the criticality of a system can be evaluated by using it, instead of modeling all fuel assemblies explicitly. To determine the nuclide composition of a spent fuel, a simple method is studied utilizing a large number of nuclide composition data stored in the database. Further, the effects of the axial burnup profile and calculation errors are discussed, and the remaining tasks are identified

Temperature Effects on Reactivity in Light Water Moderated UO2 Cores with Soluble Poisons

Experimental and computational studies have been performed on the temperature coefficients of reactivity in light-water moderated and reflected UO2 cores with soluble poisons such as boron and gadolinium. Experiments were carried out using the Tank-type Critical Assembly (TCA) in Japan Atomic Energy Research Institute (JAERI). Temperature coefficients of the cores with soluble poisons were measured by changing the temperature of the moderator and reflector from the room temperature to about 60°C. The dependence of temperature coefficients on the core configuration and the concentration of soluble poison was studied with the water level worth method. Temperature coefficients were calculated with a diffusion code CITATION included in the SRAC code system and a perturbation code CIPER for comparison with the experimental data. It was found that the temperature coefficients are always negative in the experimental cores (the water to fuel volume ratio (Vm/Vf) of 1.83) containing boron as soluble poison. On the...

Benchmark experiments of thermal neutron and capture gamma-ray distributions in concrete using 252Cf

Abstract The distributions of thermal neutrons and capture gamma rays in ordinary concrete were investigated by using 252Cf. Two subjects are considered. One is the benchmark experiments for the thermal neutron and the capture gamma-ray distributions in ordinary concrete. The thermal neutron and the capture gamma-ray distributions were measured by using gold-foil activation detectors and thermoluminescence detectors. These were compared with the simulations by using the discrete ordinates code ANISN with two different group structure types of cross-section library of a new Japanese version, JENDL-3.3, showing reasonable agreement with both fine and rough structure groups of thermal neutron energy. The other is a comparison of the simulations with two different cross-section libraries, JENDL-3.3 and ENDF/B-VI, for the deep penetration of neutrons in the concrete, showing close agreement in 0- to 100-cm-thick concrete. However, the differences in flux grow with an increase in concrete thickness, reaching up to approximately eight times near 4-m thickness.

Measurements and Analyses of Reactivity Effect of Fission Product Nuclides in Epithermal Energy Range

Experimental data usable for evaluating cross sections of main fission product elements (Rh, Cs, Nd, Sm, Eu and Gd) in the epithermal energy range were measured. A cadmium-covered vessel containing a pure water or an aqueous solution of a fission product element was inserted at the center of TCA (Tank-type Critical Assembly) core. Reactivity effects were obtained by the difference in the critical water levels between a pure water and an aqueous solution in the vessel. The measured reactivity was more than 1 φ and it was greater than the experimental uncertainties. Since the adjoint thermal flux below the cadmium-cutoff energy are largely depressed in the vessel, the reactivity effects in epithermal energy range could be measured. The analyses for the experiments were performed using the SRAC code system and neutron transport calculation code TWOTRAN. The exact Perturbation theory was applied to calculate the reactivity effects of fission product elements. The calculated reactivity effects using JENDL-3.2 ...

Measurements of Thermal Disadvantage Factors in Light-Water Moderated PuO2-UO2 and UO2 Lattices

The disadvantage factor for thermal neutrons in light-water moderated PuO2-UO2 and UO2 square lattices were obtained from measurements of thermal neutron density distributions in a unit lattice cell, measured with Dy-Al wire detectors. The lattices consisted of 3.4w/o PuO2-UO2 and 2.6w/o UO2 fuel rods, and the water-to-fuel volume ratio within the unit cell was parametrically changed. The PuO2-UO2 and UO2 fuel rods were designed to realize equal fissile atomic number density. The disadvantage factors thus measured were 1.36±0.07, 1. 37±0.08, 1.40±0.06 and 1.38±0.06 in the PuO2-UO2 fuel lattices, and 1.30±0.06, 1.31±0.08, 1.30±0.08 and 1.33±0.06 in the UO2, for water-to-fuel volume ratios, of 1.76, 2.00, 2.38 and 2.95, respectively. This difference in disadvantage factor between PuO2-UO2 and UO2 fuel lattices corresponds to about 8%. Calculated results obtained by multigroup transport code LASER agreed well with the measured ones.