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AbstractA series of integral experiments were conducted at the fast critical assembly (FCA) of the Japan Atomic Energy Agency, simulating light water reactor cores with a tight lattice cell of highly enriched mixed-oxide (MOX) fuel containing >15% fissile plutonium (Pu). The three experimental configurations of the FCA-XXII-1 assembly were constructed using foamed polystyrene with different void fractions (45%, 65%, and 95%) to clarify the prediction accuracy of neutronics calculation codes and nuclear data libraries among various neutron spectra. The hydrogen-to–nuclear fuel atomic ratio varied from 0.1 to 0.8. The nuclear characteristics measured in the experiments were criticality (keff), moderator void reactivity worths, and sample reactivity worths using boron carbide (20%, 60%, and 90% 10B enrichment) and Pu (92%, 81%, and 75% fissile Pu ratio).Preliminary analyses on experiments were conducted using a deterministic calculation code system conventionally used for fast reactors and the Japanese evaluated nuclear data library of JENDL-4.0. The calculated keffvalues overestimated the experiments beyond the experimental uncertainties. However, most reactivity worth calculations agreed well with the experimental values. Even beyond the experimental uncertainties, discrepancies between the calculation and the experiment were <13%.Specifically in the reactivity worth analyses of the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations for criticality and large reactivity worths were performed with the Monte Carlo calculation code MVP3 by modeling the experimental configurations in detail, confirming that the deterministic calculations closely agreed with the reference values.

ABSTRACTAir dose rates within a range of 80 km from the Fukushima Dai-ichi Nuclear Power Plant measured via car-borne surveys using the KURAMA systems from 2011 to 2018 were analyzed. The ratio of the air dose rates relative to the first car-borne survey was 0.08 over a period of seven years, indicating a much more rapid decrease than that estimated by the physical decay of radiocesium (0.3). Differences in the decreasing trends of the air dose rates among the land-use categories and evacuation order areas were examined. The ecological half-lives were evaluated according to the land-use categories in each evacuation order area. To model the acceleration of the decrease in the air dose rates observed in the evacuation order area, we modified the bi-exponential curve formula of ecological half-life and introduced the two-group model. The ecological half-life of the fast components in the evacuation order area of ‘zone in preparation for the lifting of the evacuation order’ evaluated using the two-group model after 2013 were 0.5–1 y, which were much shorter than those before 2013 (2–3 y).

A reflector reactivity worth was measured by replacing stainless steel with zirconium at the FCA. The experimental result of the positive reflector reactivity worth demonstrates the effectiveness of the zirconium reflector compared with the SS reflector in the fast reactor core. This paper also focuses on the validation of standard calculation methods used for fast reactors with JENDL-4.0. As a result, it is confirmed that the standard calculation methods for the reflector reactivity worth show agreement within the experimental error.

Two extended bias factor methods, the LC and PE methods, were applied to the prediction accuracy evaluation of neutronic characteristics of a breeding light water reactor, using data of FCA-XXII-1 critical experiments, in order to investigate the features and effectiveness of these methods on the basis of an actual core design and existing experimental results. The present study confirms the following features of these methods. Both the LC and PE methods can improve the prediction accuracy the most when all the experimental results are used. The prediction accuracy improvement is achieved mainly by reducing uncertainty due to errors in cross sections. This is done by realizing a profile of sensitivity coefficients closer to that of the target core and suppressing the influence of errors in experiments and experimental analysis methods. The PE method always improves the prediction accuracy with the use of any combination of experimental results. It is always superior to the LC method in the improvement of the prediction accuracy. Concerning the effectiveness of using the extended bias factor methods with the data of FCA XXII-1 critical experiments, it is concluded that the experimental results regarding multiplication factor are more effective than the other experimental results, namely, reaction rate ratios of 238U capture to 239Pu fission (C28/F49) and void reactivity, in reducing prediction uncertainties of all the neutronic characteristics of the target core investigated: the multiplication factor, the C28/F49, and the void reactivity of the target core. This is due to the fact that the extended bias factor methods cannot fully utilize the potential that these experimental results have for the reduction of the uncertainties due to the errors in cross sections because of their strong correlations to the target core characteristics. This failure is due to large errors in the experiments and/or the experimental analysis methods.

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.

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 LMFBR's, 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.

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.

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 was used, the calculation agreed with experimental values within the experimental errors (–6 a%t 2,000°C).The C/Evalues in the high temperature range were compared between the activation Doppler effect and the Doppler reactivity worth. Both values showed good agreement within the experimental errors. When the activation Doppler effect can be considered as complementary to the Doppler reactivity measurements, the calculation accuracy of the Doppler reactivity worth at 2,000°C was confirmed. A series of the high temperature Doppler effect measurements at the FCA extend experimental temperature range and contribute to the improvement of the calculation reliability of the 238U Doppler effect at higher temperature.

A study on the nuclear characteristics of an axially decoupled core was performed through a series of experiments using the solid moderator core in the Kyoto University Critical Assembly (KUCA). In the axially decoupled core which has an internal blanket between the upper and lower cores, the measured differential reactivity worth of a control rod moving axially was apparently asymmetric between the two core regions. Since it was considered that this phenomenon was attributed to the neutron flux tilt caused by the axial movement of control rods during the measurement of differential worth, the axial reaction rate distributions were measured by gold wires with and without a cadmium tube. It was found that the reaction rate distribution was remarkably tilted by changing the pattern of control rods inserted in the core to maintain the critical state, and that the flux distribution was tilted without energy dependency. The value of first-order eigenvalue separation was obtained using the relation between the neutron flux tilt and the first-mode eigenvalue separation based on the Explicit Higher-order Perturbation (EHP) method. This value agreed with the calculated result by the SRAC system.

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 based on the Benoist's theory was validated by transposing the plate direction. Moreover, the design concept of LMFBR with the sodium plenum, which aims at the reduction of the sodium void reactivity, was also examined.