Hiroyuki Oigawa

Neutronics design for lead-bismuth cooled accelerator-driven system for transmutation of minor actinide

Neutronics design study was performed for lead-bismuth cooled accelerator-driven system (ADS) to transmute minor actinides. Early study for ADS indicated two problems: a large burnup reactivity swing and a significant peaking factor. To solve these problems, effect of design parameters on neutronics characteristics were searched. The design parameters were initial plutonium loading, buffer region between spallation target and core, and zone fuel loading. Parametric survey calculations were performed considering fuel cycle consisting of burnup and recycle. The results showed that burnup reactivity swing depends on the plutonium fraction in the initial fuel loading, and the lead-bismuth buffer region and the two-zone loading were effective for solving the problems. Moreover, an optimum value for the effective multiplication factor was also evaluated using reactivity coefficients. From the result, the maximum allowable value of the effective multiplication factor for a practical ADS can be set at 0.97. Consequently, a new core concept combining the buffer region and the two-zone loading was proposed base on the results of the parametric survey.

Corrosion–erosion test of SS316 in flowing Pb–Bi

Abstract Corrosion tests of austenitic stainless tube were done under flowing Pb–Bi conditions for 3000 h at 450 °C. Specimens were 316SS produced as a tubing form with 13.8 mm outer diameter, 2 mm thickness and 40 cm length. During operation, maximum temperature, temperature difference and flow velocity of Pb–Bi at the specimen were kept at 450, 50 °C, and 1 m/s, respectively. After the test, specimen and components of the loop were cut and examined by optical microscope, SEM, EDX, WDX and X-ray diffraction. Pb–Bi adhered on the surface of the specimen even after Pb–Bi was drained out to the storage tank from the circulating loop. Results differed from a stagnant corrosion test in that the specimen surface became rough and the corrosion rate was maximally 0.1 mm/3000 h. Mass transfer from the high temperature to the lower temperature area was observed: crystals of Fe–Cr were found on the tube surface in the low-temperature region. The sizes of crystals varied from 0.1 to 0.2 mm. The depositing crystals were ferrite grains and the chemical composition ratio (mass%) of Fe to Cr was 9:1.

Experiments and analyses on sodium void reactivity worth in mock-up cores of metallic fueled and MOX fueled fast reactors at FCA

Sodium void reactivity worth in two metallic fueled fast reactors and a MOX fueled one was measured at the Fast Critical Assembly (FCA). Using the measured reactivity worth, bias factors to be applied to the calculation were determined for the non-leakage term and the leakage term separately by a least squares method. Based on these bias factors, the calculation accuracy for each term was assessed. For the non-leakage term, JENDL-3.2 showed good agreement with the experiments. No difference in the calculation accuracy was observed between the metallic fuel and the MOX fuel, though their neutron spectra were largely different. When the enriched uranium was contained in the void region together with plutonium, however, JENDL-3.2 overestimated the non-leakage term by about 6 %. The effect of the plutonium isotope composition was also examined. For the leakage term, it was found that the accuracy of the current calculation depended upon the void fraction in the cell. The directional diffusion coefficient base...

A Proposal of Benchmark Calculation on Reactor Physics for Metallic Fueled and MOX Fueled LMFBR Based upon Mock-up Experiment at FCA

In order to assess the validity of the cross section library for fast reactor physics, a set of benchmark calculation is proposed. The benchmark calculation is based upon mock-up experiments at three FCA cores with various compositions of central test regions, two of which were mock-ups of metallic fueled LMFBRs, and the other was a mock-up of a mixed oxide fueled LMFBR. One of the metallic cores included enriched uranium in the test region, while the others did not. Physics parameters to be calculated are criticality, reaction rate ratios, plutonium and B4C sample worth, sodium void reactivity worth, and Doppler reactivity worth of 238U. Homogenized atomic number densities and various correction factors are given so that anyone can easily perform diffusion calculation in two-dimensional RZ-model and compare the results with the experiments. The validity of the correction factors are proved by changing the calculation method and used nuclear data file.

Integral Validation of Minor Actinide Nuclear Data by using Samples Irradiated at Dounreay Prototype Fast Reactor

The reliability of nuclear data for minor actinides was evaluated by using the results of the post‐irradiation experiment for actinide samples irradiated at the Dounreay Prototype Fast Reactor. The burnup calculations with JENDL‐3.3, ENDF/B‐VI.8, and JEFF‐3.0 were performed. From the comparison between the experimental data and the calculational results, in general, the reliability of nuclear data for the minor actinides are at an adequate level for the conceptual design study of transmutation systems. It is, however, found that improvement of the accuracy is necessary for some nuclides, such as 238Pu, 242Pu, and 241Am.

Experiments and Analyses on Sodium Void Reactivity Worth in Uranium-Free Fast Reactor at FCA

A uranium-free fast reactor was simulated at FCA in order to examine the prediction accuracy for sodium void effect of plutonium burning fast reactors. Material sample worth for plutonium and B4C was also measured to compare its prediction accuracy with that for the sodium void reactivity worth. It was found that an axial distribution of plutonium sample worth and the central B4C sample worth for various kinds of 10B enrichment were precisely calculated by the conventional calculation method for fast reactors with 70-group structure. The sodium void reactivity worth was, however, poorly predicted especially for the non-leakage term. This discrepancy seems to be caused by the peculiar energy breakdown of the non-leakage term in the uranium-free fast reactor.

Experiments and Analyses on Fuel Expansion and Bowing Reactivity Worth in Mock-up Cores of Metallic Fueled Fast Reactors at FCA

Axial fuel expansion and radial fuel bowing were simulated in mock-up cores of metallic fueled fast reactors at the Fast Critical Assembly (FCA). Reactivity worth caused by the simulation was measured and compared with calculations. Based on these experiments and calculations, the applicability of current calculation methods was discussed for both the first order perturbation theory (FOP) and the exact perturbation theory (EP). For the axial fuel expansion reactivity worth, both FOP and EP showed 10 to 20% smaller values than the experiment. This underestimation was consistent to a C/E trend of axial distributions of plutonium sample worth. No significant difference was observed between FOP and EP, when transport correction was applied. For the radial fuel bowing reactivity worth, the FOP showed about 10% larger values than the EP. Near the core central plane, the EP with transport correction showed good agreement with the experiment, while FOP showed overestimation by 14%. At the core axial edge, however...

High temperature Doppler effect experiment for 238U at FCA, (II) : Reaction rate measurement up to 2,000°C with a foil heated by laser exposure

For evaluation and improvement of the accuracy of calculations of the Doppler effect in a temperature range up to 2,000°C, a new experimental device was developed to measure the 238U Doppler effect at the Fast Critical Assembly of Japan Atomic Energy Research Institute. The measurement technique is based on a reaction rate measurement with a heated foil. Components of the device that are subjected to high temperatures are made of tungsten, while the remaining structural parts are made of stainless steel. The measurements were carried out in a mock-up core of an oxide-fueled fast reactor, where the Doppler reactivity worths up to 1,500°C were previously measured. In the analysis, the activation Doppler effect was determined by precise calculation of the effective capture rate in the Doppler foil. A collision probability cell code with ultra-fine group structure, PEACO-X, was used in the resonance energy range and the conventional cell code SLAROM was used in all other energy ranges. When the JENDL-3.2 data...

High temperature Doppler effect experiment for 238U at FCA, (I) Reactivity worth measurement with a small heated sample up to 1,500°C

The experimental device for Doppler effect measurement was developed for the Fast Critical Assembly of Japan Atomic Energy Research Institute. The Doppler effect up to 1,500°C can be measured with this device. The experimental data is to be used for evaluating and improving the calculation method of the Doppler effect in higher temperature range. The measurement technique is based on a reactivity worth measurement with a heated sample. The device is made of tungsten for the high temperature parts and stainless steel for the other structural parts. The measurements were carried out in a mock-up core of an oxide-fueled fast reactor. In the analysis of the experimental data, the Doppler reactivity worth was calculated by a first order perturbation theory and JENDL-3.2 library. A collision probability cell code with ultra-fine group structure, PEACO-X, was used to obtain the precise effective cross sections in a pin cell model of the Doppler sample. The calculation underestimates Doppler effect by about 5% in high temperature range. The Doppler effect above 1,500°C was estimated by the extrapolation of the measured data below this temperature. When the calculated and extrapolated Doppler reactivity worths at 2,000°C were compared, a good agreement between them was obtained. The significant improvement of calculation reliability of the Doppler effect in the higher temperature range was achieved by this experiment.

Shielding Analysis of the Secondary Coolant Circuit of Accelerator Driven Systems

Disposition of radioactive waste is one of the key issues to make nuclear energy as a sustainable power resource for future power generation. To solve this issue, a long-term program for research and development on partitioning and transmutation called “OMEGA” (Options Making Extra Gains from Actinides and fission products) was adopted in 1988 by the Japan Atomic Energy Commission (Japan Atomic Energy Commission, 2009). The aims of the OMEGA Program are to widen options for future waste management and to explore the possibility to utilize high-level radioactive wastes as useful resources. To proceed the OMEGA program, partitioning and transmutation (P-T) is a key technology to reduce environmental impacts and source terms contained in the high level radioactive waste which was discharged from commercial nuclear power reactors. By using partitioning technology, the long-lived radioactive nuclides such as minor actinides (MAs) and longlived fission products (LLFPs) can be extracted from high level radioactive waste. From the mass balance study performed under the OMEGA program, the lifetime of final disposal site can be prolonged 3 to 10 times more than current designed lifetime (Oigawa, et al. 2005). Under the framework of the OMEGA program, Japan Atomic Energy Agency (JAEA) has proposed a double-strata fuel cycle concept (Takano, et al. 2000) which consists of two fuel cycles; one is a commercial fuel cycle including current LWR cycle and also future FBR fuel cycle and the other is a small fuel cycle dedicated to the transmutation of MA and LLFP. By the double-strata fuel cycle shown in Fig. 1, long-lived radioactive nuclides are confined into the second-stratum small fuel cycle that includes an innovative nuclear system which is optimized to the transmutation of MA and LLFP. JAEA carries out research and development of accelerator-driven transmutation systems (ADS) as an innovative dedicated transmutation system under the OMEGA Program. In the P-T, the most effective reaction to transmute MA separated from high level radioactive waste by partitioning process into short-lived/stable nuclides is fission reaction. By fission reaction, target MA nuclides can be directly transmuted into short or stable nuclides. However, to make fission reaction as dominant reaction in the nuclear reactor, the reactor core should be formed a fast neutron spectrum field. It requires rather large fuel inventory and diminishes some safety characteristics such as Doppler coefficient and delayed neutron fraction. Another point of view, MA discharged from current light water power reactors is roughly composed of 55% of neptunium, 40% of americium and 5% of