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In a severe accident of a lead–bismuth-cooled accelerator-driven system, a gas–liquid two-phase flow with a large liquid-to-gas density ratio might appear, such as a steam leakage into hot lead–bismuth flow. It is still difficult to predict such phenomena because there are no available flow models for two-phase flow with a large density ratio compared to ordinary two-phase flows such as an air–water two-phase flow. Therefore, a two-phase flow model should be developed based on experimental data of two-phase flows with a large density ratio. In this study, a liquid–metal two-phase flow was measured by using a four-sensor electrical conductivity probe and a miniature electromagnetic probe to establish an experimental database for lead–bismuth flow structure. In measurements with the four-sensor probe, the radial profiles of void fraction and interfacial area concentration were measured at different axial positions. Experiments were also performed to understand the turbulent structure in a liquid–metal two-phase flow by using the electromagnetic probe. From the data measured by both four-sensor and electromagnetic probes, it is shown that the turbulence intensity at the pipe center was proportional to the void fraction to the power of 0.8 for higher void fraction. These results represented a similar tendency as previous data in air–water two-phase flows.

This chapter discusses the theory of higher-order modes in the Feynman function and cross-power spectral density (CPSD) in an accelerator-driven system (ADS) where pulsed spallation neutrons are injected at a constant time interval. Theoretical formulae that consider the higher-order modes of the correlated and uncorrelated components in the Feynman function and CPSD for an ADS were recently derived in a paper published by the author. These formulae for the Feynman function and CPSD are applied to a subcritical multiplying system with a one-dimensional infinite slab geometry in this chapter. The Feynman functions and CPSD calculated with the theoretical formulae are compared with the Monte Carlo simulations of these noise techniques. The theoretical formulae reproduce the Monte Carlo simulations very well, thereby substantiating the theoretical formulae derived in this chapter. The correlated and uncorrelated components of the Feynman functions and CPSD are decomposed into the sum of the fundamental mode and higher-order modes. This chapter discusses the effect of subcriticality on the higher-order mode effects.

The neutron spectrum of a pulsed neutron reactor at subcritical state is different from that evaluated by -eigenvalue mode, because of the time needed in the neutron slowing-down process from fast to thermal energy range. The time needed in slowing down does not depend on the degree of subcriticality, but the decreasing speed of neutron flux becomes fast as the subcriticality becomes deep. Therefore, the neutron spectrum becomes soft as the subcriticality becomes deep. This fact suggests to us that group constants to be used in the design study should change with the degree of subcriticality of the target system, even in the case of the same composition.

The RBWR (resource-renewable boiling water reactor) is an innovative BWR that has a capability to breed and burn trans-uranium elements (TRUs) using a multi-recycling process. The RBWR can be used as a long-term energy supply, and it reduces the negative environmental impact that TRUs cause as they are otherwise long-lived radioactive wastes. Various design concepts of the RBWR core have been proposed. The RBWR-AC is a break-even reactor and the RBWR-TB and RBWR-TB2 are TRU burners. The RBWR-TB is designed to burn TRUs from the RBWR-TB itself and to burn almost all the TRUs by repeating their recycling. The RBWR-TB is assumed to be applied for a nuclear power phase-out scenario. The RBWR-TB2 is intended to burn TRUs from LWR spent fuels. The RBWR-TB2 is assumed to be applied for reducing the amount of TRUs to be managed in storage facilities. The RBWR cores achieve their TRU multi-recycling capability under the constraint that the void reactivity coefficient must be negative by introducing the parfait core concept. This chapter reviews details of the specific design and core characteristics of the RBWR.

A TRU-burning fast reactor cycle associated with a uranium-free trans-uranium (TRU) metallic fuel core is one of the solutions for radioactive waste management issue. Use of TRU metallic fuel without uranium makes it possible to maximize the TRU transmutation rate in comparison with uranium and plutonium mixed-oxide fuel because it prevents the fuel itself from producing new plutonium and minor actinides, and furthermore because metallic fuel has much smaller capture-to-fission ratios of TRU than those of mixed-oxide fuel. Also, adoption of metallic fuel enables recycling system to be less challenging, even for uranium-free fuel, because a conventional scheme of fuel recycling by electrorefining and injection casting is applicable.

There are some issues, however, associated with a uranium-free TRU metallic fuel core: decrease in negative Doppler reactivity coefficient from the absence of uranium-238, which has the ability to absorb neutrons at elevated temperatures, increase in burn-up swing, because fissile decreases monotonically in uranium-free core, and so on. The purpose of this paper is to evaluate the feasibility of the uranium-free TRU metallic fuel core by investigating the effect of measures taken to enhance Doppler reactivity feedback and to reduce burn-up swing. The results show a TRU-burning fast reactor cycle using uranium-free TRU metallic fuel is viable from the aforementioned points of view because the introduction of diluent Zr alloy, spectrum moderator BeO, and lower core height enables Doppler reactivity coefficient and burn-up reactivity swing of uranium-free TRU metallic fuel to be as practicable as those of conventional fuel containing uranium.

A hydride target including minor actinides (MA) is able to enhance the transmutation rate in a fast breeding reactor (FBR) without degradation of core safety characters. Fast neutrons generated in the core region of the FBR are moderated in the MA-hydride target assemblies and then efficiently absorbed by MA. The MA-hydride target pin has been designed in the light of recent research of hydride materials. This chapter shows the feasibility of MA transmutation by an existing reactor, Monju.

To effectively transmute minor actinides (MAs), which have long-lived radioactivity and high decay heat, fast reactors are very promising because many minor actinides can be loaded and transmutation rates are high compared to light water reactors. With the increase of loaded minor actinides, the neutron spectrum becomes hard and core safety parameters will deteriorate. Especially, the sodium void reactivity increases with MA addition to cores. To overcome the difficulty, we propose MA transmutation fast reactors using core concepts with a sodium plenum and internal blanket region in reactor cores. Therefore, cores become complex, and calculation accuracy becomes poor. To accurately evaluate the neutronic properties such as MA transmutation rate and sodium void reactivity, we improved calculation methods. In this chapter we show new methods for calculating MA transmutation rates for each MA nuclide, for calculating the uncertainty of MA transmutation using sensitivities. A new sensitivity is derived that is defined as a relative change of core parameters relative to infinite-dilution cross sections, not effective cross sections. To eliminate bias factors in estimating core parameter uncertainties, a new method is proposed. This method is used to reduce the calculation uncertainty through the use of adjusted cross sections.

Since the early 1970s, studies and experimental projects have been undertaken in Europe to examine the potential of thorium-based fuels in a variety of reactor types. The first trials were mainly devoted to the use of thorium in high-temperature reactors. These projects can be seen as scientific successes but were not pursued on a commercial basis because of the priority given in Europe to the development of light water reactors. Later on, thorium oxide was considered as a potential matrix for burning plutonium (possibly also minor actinides), and several core design studies, as well as experiments, were undertaken. The most recent such concern the BR2 and HFR Material Test Reactor (MTR) irradiations in Belgium and in the Netherlands, respectively, as well as the KWO PWR in Obrigheim in Germany, in which thorium-plutonium oxide fuel (Th-MOX) was successfully irradiated up to 38 GWd/tHM. The results of these experiments have shown that Th-MOX behaves in a comparable way as conventional uranium-plutonium oxide fuel (U-MOX). More work is still needed before Th-MOX will reach sufficient maturity to implement it on a large scale in power reactors, but all currently available results indicate that licensing Th-MOX for LWRs should be feasible. Finally, European research projects are still devoted to the study of thorium salts in molten salt reactors, a design that incorporates on-line reprocessing and needs no specific thorium fabrication, adding therefore the benefits of thorium without its main challenges.

With consideration of the phase-out option from nuclear power (NP) utilization in Japan, an accelerator-driven system (ADS) for Pu transmutation has been designed and scenario analysis performed. The ADS is designed based on the existing ADS design for MA transmutation, and the six-batch ADS was selected as a reference design for scenario analysis. In the scenario analysis, the once-through scenario of light water reactor (LWR) spent fuel is referred to as a conventional scenario with a LWR-MOX utilization scenario. As the transmutation scenario, three cases of transmuters that are only-FR, only-ADS, and both-FR+ADS are analyzed. The numbers of necessary transmuters are obtained as 15 to 32 units, and the necessary period for transmutation as 180–240 years. The benefit on repository by reduction of Pu and MA is reduction of repository area by a factor of five and of decay time of toxicity by one order of magnitude. The FR+ADS scenario would be a modest solution, although the ADS scenario is preferable if rapid transmutation is required.

Sensitivity analyses of initial compositions and cross sections were conducted to quantitatively clarify the source elements and the nuclear reactions dominating the generation of activation products. In these analyses, the ORIGEN2.2 code was used with ORLIBJ40, a set of the cross-section libraries based on JENDL-4.0. Analyses were conducted for the activations of cladding tubes, end plugs, and spacers of fuel assemblies and channel boxes in BWR that are composed of zirconium alloy, stainless steel, and nickel-chromium-based alloy. From about 50 representative radioactive nuclides, several nuclides were selected as the targets of sensitivity analyses for the aspect of their large concentrations in the target materials.

The results of sensitivity coefficients clarified the source elements and the nuclear reactions dominating the generation of activation products even for the nuclides generated through complicated pathways. These results could be utilized to select the objectives of the impurity elements for measurements and of nuclear data for the improvement of accuracy. These results will contribute to improvements in the accuracy of numerical evaluations of activation product concentrations.