Gunzo Uchiyama

Lattice physics analysis of measured isotopic compositions of irradiated BWR 9 × 9 UO2 fuel

As part of a validation study of burnup calculations of BWR cores, lattice physics analyses were performed on burnups and isotopic compositions of U, Pu and fission product nuclides measured on five samples taken from 9 × 9 BWR fuel assemblies. Burnup calculations in infinite assembly geometry were carried out using MVP-BURN and SRAC codes coupled with major nuclear data libraries. The burnups determined based on the Nd-148 method were from 27.9 to 64.2 GWd/t. The typical relative differences in isotopic compositions (atom/Total-U) between the burnup calculations and measurements were −2 ∼ 19% for 234U, −20 ∼ 3% for 235U, −1.5 ∼ 0.1% for 236U, −0.04 ∼ 0.02% for 238U, −4 ∼ 11% for 238Pu, −11 ∼ −2% for 239Pu, −3 ∼ 0% for 240Pu, −12 ∼ −2% for 241Pu and −2 ∼ 3% for 242Pu. They were −2 ∼ 2% for Nd isotopes, −15 ∼ 7% for Eu isotopes, −13 ∼ 1% for Cs isotopes, −13 ∼ 8% for Sm isotopes, 0 ∼ 7% for 147Pm, −7 ∼ −2% for 95Mo, −2 ∼ −1% for101Ru and 0 ∼ 4% for 103Rh.

Measurement of Neutron Dose under Criticality Accident Conditions at TRACY Using TLDs

Applicability of a neutron dose equivalent monitor with thermoluminescence dosimeter (TLD monitor) as an area dosimeter for criticality accidents was studied through measurements at Transient Experiment Critical Facility (TRACY). The TLD monitor is composed of two TLD badges and a cubical polyethylene case, and has a response similar to the values of the conversion coefficients for the ambient dose equivalent in the energy range from 0.1 to 10MeV. TRACY was operated with and without a water reflector when irradiating the TLD monitors. The neutron doses measured with the TLD monitors were converted into tissue kerma using dose conversion factors calculated with MCNP5. Factors to correct for the difference between responses of the TLD monitor to the spontaneous fission spectrum of a 252Cf calibration source and to spectra of TRACY were also calculated with MCNP5 and applied to the tissue kerma. The values of tissue kerma were proportional to the integrated power of TRACY. The measured tissue kerma ranged from 0.034 to 16 Gy, which covers the range from 0.1 to 10Gy specified as important in criticality accident dosimetry by the IAEA. The presented method of measurement also satisfies the time limit on dose determination.

Benchmark Critical Experiments and FP Worth Evaluation for a Heterogeneous System of Uranium Fuel Rods and Uranium Solution Poisoned with Pseudo-Fission-Product Elements

A series of critical experiments were performed using heterogeneous cores at the Static Experiment Critical Facility (STACY) of Japan Atomic Energy Agency (JAEA) in order to obtain systematic benchmark data concerning the dissolving process in a reprocessing plant. Focusing on the introduction of the burn-up credit, critical mass measurement was conducted for a combination of uranium dioxide fuel rods (5 wt% 235U) and uranyl nitrate solution (6 wt% 235U) poisoned with pseudo-fission-product (FP) elements—samarium, cesium, rhodium, and europium. Fuel rods were arrayed at a 1.5-cm lattice interval in the poisoned fuel solution in a 60-cm-diameter cylindrical tank. The uranium concentration of the solution was roughly kept at about 320 gU/L, and the FP element concentrations were adjusted to be equivalent to that in a burn-up of about 30GWd/t. The result provides basic experimental data for validation of computational methods to evaluate the reactivity effect of each FP element, as well as benchmark criticality data for validation of neutron multiplication factor calculation for heterogeneous systems of spent fuel. In this report, the details of the experiments and benchmark models will be presented as well as the procedure and the result of separate reactivity worth evaluation for each FP element. The experimental results and the computational evaluation results will also be compared.

Hydrogen Absorption Behavior of Titanium Alloys by Cathodic Polarization

Titanium and Ti-5mass%Ta alloy has been utilized in nuclear fuel reprocessing plant material because of its superior corrosion resistance in nitric acid solutions. However, Ti alloy have been known to high susceptibility of hydrogen embrittlement. To evaluate properties of hydrogen absorption and hydrogen embrittlement of Ti alloys, cathodic polarization tests and slow strain rate tests (SSRT) under cathodic polarization were carried out. Results show titanium hydrides covered on the surface of metals and hydrides thickness were within 10μm. But hydride did not observed at inner part of metals. Ti and Ti-5%Ta did not show hydrogen embrittlement by SSRT under cathodic charging. These results suggested that Ti and Ti-5%Ta could absorb hydrogen. But hydrogen did not penetrate inner portion of the metals more than 10μm in depth because titanium hydrides act as barrier of hydrogen diffusion. It is considered that retardation of hydrogen diffusion hindered hydrogen embrittlement of Ti and Ti-5%Ta alloys.Copyright

Corrosion Study of Titanium-5% Tantalum Alloy in Hot Nitric Acid Condensate

In this study, a long-term corrosion tendency and metal salt effect in heating nitric acid solution on corrosion behavior of titanium-5% tantalum alloy (Ti-5Ta) in hot nitric acid condensate condition were mainly researched to discuss the aging behavior of reprocessing equipments such as evaporators made of titanium or its alloy. The hot pure nitric acid solution with continuous renewing such as the nitric acid condensate condition is severe corrosion environment for their materials because of the corrosion inhibition effect from titanium ions as corrosion products or oxidizing ions in nitric acid solution and is certainly formed in evaporator for spent nuclear fuel reprocessing. From the results of the long-term corrosion test for total 11,000 hrs, the corrosion of Ti-5Ta in the nitric acid condensate was accelerated with increase of the nitric acid concentration in the condensate (∼5.6 M). The corrosion rate was nearly constant during the immersion time and the test coupons suffered a uniform corrosion. Thus, from the viewpoints of nitric acid corrosion, the life-time of the reprocessing equipments made of titanium or its alloy will be roughly estimated based on the results of average corrosion rate in operation. It was also found that the kind and concentration of metal salt in the heating nitric acid solution gave a remarkable effect on the concentration of nitric acid vapor and the corrosion of Ti-5Ta in the hot nitric acid condensate. Most of the evaporators for reprocessing plants include metal ions in the heating nitric acid solution, so the metal salt effect is one of the corrosion factors to control the corrosion behavior of titanium alloy in condensate. The nitric acid concentration in the condensate increases by adding the metal salts in the heating nitric acid solution, in addition, the larger valence of metal ions was contributed to the increase of nitric acid concentration in the condensate. Consequently, the metal salts effect in the heating nitric acid solution accelerates the corrosion of Ti-5Ta in the nitric acid condensate. The corrosion of titanium or its alloy in nitric acid condensate condition should be carefully considered as one of severe corrosion environment in evaporators for reprocessing plant. This corrosion study would give useful information to estimate the lifetime of evaporators made of titanium alloy from the viewpoint of nitric acid corrosion.Copyright

Study on Corrosion of Stainless Steel in Boiling Nitric Acid Under Heat Transfer Conditions

In PUREX process for spent fuel reprocessing plants, nitric acid solutions are treated in many components. Particularly, heating portions in the components are severely corroded in the boiling solution under heat transfer (HT) conditions. Although corrosion behavior under the conditions has been investigated by many researchers, corrosion mechanism has not been sufficiently understood. Consequently, because of its safety, we need to study dominant factors of corrosion mechanism and to develop a method to predict corrosion progress.Surface temperature, heat flux, concentration and compositions of the solution have been previously considered as important factors for corrosion rate (CR). In this paper, authors have focused on the effects of surface temperature and heat flux on CRs of stainless steels in boiling nitric acid under HT conditions. We performed experimental study in consideration of dissolvers and concentrators. Two types of cells for HT and immersion conditions were applied for corrosion tests and the effects were compared. Test solution used was 33 mol·m−3 (0.033 M) vanadium added to 3 kmol·m−3 (3 M) nitric acid solution. The solution was heated at boiling temperature under atmospheric pressure. Additionally, a boiling curve which was indicated the relation between heat flux and degree of superheating was investigated experimentally. Surface temperatures during corrosion tests were estimated from a boiling curve. The relationship among measured CRs, surface temperature and heat flux were studied.The results showed that CR did not depend on heat flux, but depended on surface temperature. Arrhenius plots on CRs indicated that CR was accelerated by solution boiling against non-boiling. The cause of the acceleration of CR under boiling nitric acid solution was discussed.Copyright

Benchmark Critical Experiments of a Heterogeneous System of Uranium Fuel Rods and Uranium Solution Poisoned with Gadolinium, and Application of Their Results to JACS Validation

A series of critical experiments were performed using heterogeneous cores at the Static Experiment Critical Facility (STACY) of the Japan Atomic Energy Agency (JAEA) in order to obtain systematic benchmark criticality data concerning the dissolving process in reprocessing plants. Focusing on the use of gadolinium as a soluble poison, critical mass was measured for a combination of uranium dioxide fuel rods (5 wt% 235U) and uranyl nitrate solution (6 wt% 235U) poisoned with gadolinium (Gd). Fuel rods were arrayed with a 1.5 cm lattice interval in the poisoned fuel solution in a 60-cm-diameter cylindrical tank. The uranium concentration of the solution was roughly kept at about 330 gU/L, and the Gd concentrations were varied up to 0.1 gGd/L. The other series of experiments were also conducted, as reference cases, varying uranium concentration in the fuel solution without Gd. The results provided benchmark criticality data for validation of neutron multiplication factor calculation on heterogeneous systems such as a dissolver. Validation calculation of JACS based on the newly obtained benchmarks supports the justification of its utilization for the criticality safety analysis.

Evaluation of personal dosimeters and electronic modules under high-dose field

Described in this report are evaluation tests of neutron-induced malfunction of real-time personal dosimeters and electronic modules using the Transient Experiments Criticality Facility, TRACY, at the Japan Atomic Energy Agency. In the event of radiological or nuclear terrorism, radiation levels are expected to be high, especially if nuclear-critical devices are used. In such devices, fission chain reactions develop that may lead to nuclear criticality with strong lethal emissions of both neutrons and gamma-rays. A wide area of radius 400m at least should be cordoned off. In such a situation real-time dosimeters with wireless data transmission capability would provide essential support for first-response teams. The reliability of these apparatuses, however, has not been evaluated sufficiently under such high-dose conditions. We have evaluated these at TRACY, which is a reactor that can realize controlled nuclear excursions using 10% 235U-enriched uranyl nitrate solutions as fuel.

Fluctuation of the Neutron Multiplication Factor Induced by an Oscillation of the Fuel Solution System

From the viewpoint of nuclear criticality safety, it is important to comprehend the reactivity of fuel solutions induced by oscillatory movements such as earthquakes. This paper intends to figure out the reactivity of a fuel solution system with a free surface formed by oscillation by evaluating the fluctuation of the neutron multiplication factor (k eff ) obtained from a static calculation. To fulfill this intension, criticality calculations with reflecting fluid calculation results have been carried out. In the fluid calculations, the finite volume method and the volume of fluid (VOF) method have been applied in tracking the free surface formed by oscillation. The continuous energy Monte Carlo calculation method has been applied in the criticality calculations. As a result, it has been found that the variation patterns of the k eff and those of the shape of fuel solutions are classified according to oscillation frequency and the ratio of solution height to the width of the tank (H/L). If a sloshing motion is generated, the k eff fluctuates widely and has a threshold, with which we can classify the fluctuation type of the k eff , despite the kind of reflector. If H/L is above the threshold, i.e., H/L =0.4, the k eff fluctuates to a value below that obtained in the resting state. On the contrary, if H/L is below the threshold, the k eff fluctuates to a value above that obtained in the resting state. This result implies the criticality calculation for a fuel solution with a free surface using the Monte Carlo method may give a slightly smaller threshold than using other approaches.

Nuclear Criticality Safety Evaluation of a Mixture of MOX, UO2 and Additive in the Most Conservative Concentration Distribution

The nuclear criticality safety evaluation of blenders that are used at a mixed uranium-plutonium oxide (MOX) fuel plant must take into account the nonuniform distribution of powders in three principal components, i.e., MOX, uranium dioxide (UO2) and zinc stearate, which is a fuel additive. The model blender considered in this article contained a mixture of 33 wt% PUO2-enriched MOX, depleted UO2 and zinc stearate in the form of an upside-down truncated cone that was surrounded by 30-cm-thick polyethylene. To limit the number of calculation cases, the fissile plutonium mass of the mixture was fixed at 98 kg, and the total concentration of MOX and UO2 was fixed at 4.0 g/cm3. The most conservative fuel distribution with respect to nuclear criticality safety under these constraints was calculated with a two-dimensional optimum fuel distribution code OPT-TWO, so that the importance distribution of MOX and that of zinc stearate could be individually flattened by conserving the mass of each component. The OPT-TWO calculation was followed by a criticality calculation performed with the MCNP code to obtain the neutron multiplication factor of the fuel system in the optimum fuel distribution. The most conservative fuel distribution obtained in this research was typically depicted as a layer of zinc stearate embedded within the central MOX region that was surrounded by the peripheral UO2 region. An increase of up to 25% in the neutron multiplication factor was found; two factors with comparable but independent contributions were the nonuniform concentrations of plutonium enrichment and zinc stearate.