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A photo-neutron (γ, n) reaction with laser Compton scattering γ-rays has been suggested to be effective for the nuclear transmutations of fission products. The photo-neutron reaction occurs via a giant dipole resonance, which has a large cross section and whose properties are smooth functions of mass number. The laser Compton scattering can generate effectively and selectively high-energy photons with a desired energy range. In this chapter, we investigate quantitatively the effectiveness of the transmutation with laser Compton scattering based on the Hauser–Feshbach theory using the TALYS code. We carry out simulations for high-decay heating nuclide Cs, in which the cross sections for Cs (γ, γ), (γ, n), and (γ, 2n) reactions, and the total photonuclear reaction cross sections versus incident photon energy, are calculated. The incident photon energy obtained by laser Compton scattering is also optimized. It is shown that the transmutation with medium-energy photon with a flux of more than 10/s effectively reduces the radioactivity of the target Cs.

To quantify special nuclear materials (SNM) in particle-like debris, a technique named neutron resonance densitometry (NRD) has been proposed. This method is a combination of neutron resonance transmission analysis (NRTA) and neutron resonance capture analysis (NRCA) or prompt gamma-ray analysis (PGA). In NRTA, neutron transmission rate is measured as a function of neutron energy with a short flight path time-of-flight (TOF) system. Characteristic neutron transmission dips of Pu and U isotopes are used for their quantification. Materials in the samples (H, B, Cl, Fe, etc.) are measured by the NRCA/PGA method. For the NRD measurements, a compact TOF facility is designed. The statistical uncertainties of the obtained quantities of the SNMs in a sample are estimated. A high-energy-resolution and high-S/N γ-ray spectrometer is under development for NRCA/PGA. Experimental studies of systematic uncertainties concerning the sample properties, such as thickness and uniformity, are in progress at the TOF facility GELINA of European Commission (EC), Joint Research Centre (JRC), Institute for Reference Materials and Measurements (IRMM).

We have proposed a new concept of the “self-indication method” combined with neutron resonance densitometry (NRD) for nondestructive assaying of the distribution of nuclear materials in the fuel debris of Fukushima Daiichi NPP. To verify the method, we performed experiments using a 46 MeV electron linear accelerator at the Kyoto University Research Reactor Institute. First, we measured the area densities of gold foil 10, 20, 30, 40, and 50 μm thick by area analysis at the 4.9 eV resonance region. It was confirmed that the area densities of the target nuclide can be determined by conventional NRD and the self-indication method within 3 % accuracy, respectively. As the next step, we added a silver foil of 50 μm thickness to a gold foil of 10 μm thickness and measured the area density of the gold foil. It was shown that the contribution from the other nuclide (silver foil) can be remarkably suppressed by applying the self-indication method. Finally, we have demonstrated a nondestructive assay of nuclear material using a mixture composed of a natural uranium foil, sealed minor actinide samples of Np and Am. The results indicated that the self-indication method is useful for assaying a mixture of materials with high activity such as fuel debris.

To perform decommissioning of the Fukushima Daiichi NPP safely, it is very important to measure the components of the fuel debris. Therefore, a new nondestructive assay to identify and quantify the target nuclide in fuel debris using a pulsed-neutron source is under development in Kyoto University Research Reactor Institute.

We use the self-indication method for the nondestructive assay. This method is a neutron transmission method. The neutron transmission method is focused on resonance reactions (i.e., capture, fission) at the target nuclide. In the self-indication method, the transmitted neutrons from the sample are injected into an indicator. The indicator consists of a high-purity target nuclide. The transmitted neutrons are obtained by the time-of-flight (TOF) technique via resonance reactions in the indicator. The self-indication method has a high signal-to-noise (S/N) ratio compared to the conventional method.

In this study, numerical validation for the self-indication method to identify and quantify nuclides in a BWR-MOX pellet is described. The burn-up of the MOX pellet is 0 GWd/t, 10 GWd/t, 20 GWd/t, 30 GWd/t, 40 GWd/t, and 50 GWd/t. The 12-m measurement line in KUR-LINAC is simulated as a calculational geometry. Numerical calculations are carried out by continuous-energy Monte-Carlo code MVP2 with JENDL-4.0 as the nuclear data library. The burn-up calculations of the BWR-MOX pellet are performed by the deterministic neutronics code SARC 2006 with JENDL-4.0.

Numerical validation for application of the self-indication method is carried out. From the results, it is noted that the self-indication method has a good S/N ratio compared to the neutron transmission method for quantifying the amount of target nuclides in the fuel debris.

To evaluate the feasibility of development of nuclear transmutation technology and an advanced nuclear system, precise nuclear data of neutron capture cross sections for long-lived fission products (LLFPs) and minor actinides (MAs) are indispensable. In this chapter, we present our research activities for the measurements of neutron capture cross sections for LLFPs and MAs.

In fission products in used nuclear fuel, there are several stable isotopes that have a large neutron absorption effect. For evaluation of the neutronics characteristics of a nuclear reactor, the amount of such isotopes should be evaluated by using burn-up calculation codes. To confirm the correctness of such data obtained by calculation codes, it is important to assure the precision of the evaluation of the neutron multiplication factor of used nuclear fuel. However, it is known that there are several hardly measurable elements in such important fission products. Data for the amounts of the hardly measurable elements in used nuclear fuel are scarce worldwide.

The Japan Atomic Energy Agency (JAEA) had been developing a method to assess the amounts of these fission products that are hardly measurable and have a large neutron capture cross section, under the auspices of the Japan Nuclear Energy Safety Organization. In this work, a measurement method was developed combining a simple and effective chemical separation scheme of fission products from used nuclear fuel and an inductively coupled plasma mass spectrometry with high sensitivity and high precision. This method was applied to the measurement program for the used BWR 9 × 9 fuel assembly. This measurement method is applicable to the required measurements for countermeasures to the accident at the Fukushima Dai-ichi Nuclear Power Plant of Tokyo Electric Power Company (TEPCO). JAEA has a measurement plan for not only BWR but also PWR fuel.

This presentation describes the measurement method developed in the study as well as the future measurement plan in JAEA.

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is an experimental accelerator-driven system (ADS) currently being developed at SCK•CEN for replacement of material testing reactor BR2. The MYRRHA facility is conceived as a flexible fast-spectrum irradiation facility that is able to run in both subcritical and critical modes. The applications catalogue of MYRRHA includes fuel developments for innovative reactor systems, material developments for GEN IV systems and fusion reactors, doped silicon production, radioisotope production, and fundamental science applications, thanks to the high-power proton accelerator. Next to these applications, MYRRHA will demonstrate the ADS full concept by coupling a high-power proton accelerator, a multi-megawatts spallation target, and a subcritical reactor at reasonable power level to allow operational feedback, scalable to an industrial demonstrator, and to allow the study of efficient transmutation of high-level nuclear waste. Because MYRRHA is based on the heavy liquid metal technology, namely lead–bismuth eutectic (LBE), it will be able to significantly contribute to the development of Lead Fast Reactor (LFR) technology and will have the role of European Technology Pilot Plant in the roadmap for LFR. The current design of the MYRRHA ADS and its ability to contribute to the European Commission strategy for high-level waste management through Partitioning and Transmutation (P&T) are discussed in this chapter.

After the Fukushima accident caused by the Great East Japan Earthquake, nuclear transmutation acquired much interest as an effective option of nuclear waste management. The Japan Atomic Energy Agency (JAEA) proposes the transmutation of minor actinides by an accelerator-driven system (ADS) using lead–bismuth eutectic alloy (Pb-Bi) as a spallation target and a coolant of the subcritical core. The current ADS design has 800 MWth of rated power, which is driven by a 20 MW proton LINAC, to transmute minor actinides generated from 10 units of standard light water reactors.

To obtain the data required for ADS design, including the European MYRRHA project, JAEA plans to build a Transmutation Experimental Facility (TEF) within the framework of the J-PARC project. TEF consists of two buildings: one is an ADS target test facility (TEF-T), in which will be installed a high-power Pb-Bi spallation target, and the other is the Transmutation Physics Experimental Facility (TEF-P), which will set up a fast critical/subcritical assembly driven by a low-power proton beam. TEF will be located at the end of the 400 MeV LINAC of J-PARC and accept a 250-kW proton beam with repetition rate of 25 Hz. As major research and development items of TEF-T, irradiation tests for structural materials and engineering tests for Pb-Bi applications to determine the effective lifetime of the proton beam window will be performed. The reference design parameter, that considers operating conditions of the ADS transmutor, was determined by thermal-hydraulic analyses and structural analyses. When the target operates with full-power beam, a fast neutron spectrum field is formed around the target, and it is possible to apply multipurpose usage. Various research plans have been proposed, and layout of the experimental hall surrounding the target is under way. Basic physics application such as measurements of nuclear reaction data is considered as one of the major purposes.

Experimental studies on the uranium- and thorium-loaded accelerator-driven system (ADS) are being conducted for basic research of nuclear transmutation analyses with the combined use of the core at the Kyoto University Critical Assembly (KUCA) and the fixed-field alternating gradient (FFAG; 100 MeV protons) accelerator in the Kyoto University Research Reactor Institute. The ADS experiments with 100 MeV protons were carried out to investigate the neutronic characteristics of ADS, and the static and kinetic parameters were accurately analyzed through both the measurements and the Monte Carlo simulations of reactor physics parameters. An upcoming ADS at KUCA could be composed of highly enriched uranium fuel and Pb-Bi material, and the reaction rate ratio analyses (Np and Am) of nuclear transmutation could be conducted in the ADS (hard spectrum core) at KUCA. The neutronic characteristics of Pb-Bi are expected to be examined through reactor physics experiments at KUCA with the use of solid Pb-Bi materials at the target and in the core.

In relationship to a solid target cooling system of an accelerator-driven system (ADS), wettability effect on boiling heat transfer has been experimentally investigated by irradiation with ultraviolet and gamma rays (γ-rays). The experimental apparatus consists of a copper heater block, a rectangular container, and a thermostat bath. Two copper heater blocks were fabricated: one is for radiation-induced surface activation (RISA) and the other is for photoelectric reaction by ultraviolet whose heat transfer surface is coated by a TiO film. These copper heater blocks were irradiated by ultraviolet or by γ-rays to change the surface wettability. Boiling heat transfer under subcooling conditions was measured before and after the irradiations to study the wettability effect. Experimental results show that nucleate boiling curves are shifted to the higher wall superheated side with the irradiated surface because of the decrease of the active nucleation sites. Heat transfer enhancement was found in both the critical heat flux and microbubble emission boiling (MEB) regions under these experimental conditions.