10.3.1 Experimental Setup and Procedure

A schematic view of the experimental apparatus for pool boiling experiments with small heat transfer area is shown in Fig. 10.5. The apparatus consists of a copper block, a furnace, a rectangular container, and a heat exchanger. The copper block has a cylindrical part (10 mm in diameter, 15 mm in height). Several cartridge

c

Fig. 10.4 Change of contact angle from ultraviolet, y-ray, and proton beam irradiation to TiO2 sample. (a) Ultraviolet irradiation. (b) y-Ray irradiation. (c) Proton beam irradiation heaters are installed in the copper block (3 kW in total). Two different copper blocks were fabricated to investigate the wettability effect on the heat transfer: one is a test section for ultraviolet irradiation and the other is for y-ray irradiation. Both test sections are illustrated in Fig. 10.5: three thermocouples are inserted into the cylindrical part of each copper block to estimate heat flux and surface temperature.

Copper

Distance from the surface [mm]

Thermocouple TC1*»■ TC2:6’29’ TC3* 31 <Cu> Fig. 10.5 Schematic of experimental apparatus

The heat transfer surface for ultraviolet experiments is coated by a TiO2 thin film by a sputtering process in Kyushu University. The surface for y-ray experiments is polished with #400 emery paper and then heated in air at 200 ° C for 1 h. After the thermal oxidation of the copper surface, it is irradiated in the 60Co facility where the integrated dose is 220 kGy. The irradiation conditions for heat-transfer experiments are summarized in Table 10.1.

Kenya Suyama, Gunzo Uchiyama, Hiroyuki Fukaya, Miki Umeda, Toru Yamamoto, and Motomu Suzuki

Abstract 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 x 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.

K. Suyama (*) • G. Uchiyama • H. Fukaya • M. Umeda

Japan Atomic Energy Agency, 2-4 Shirakata-Shirane, Tokai-mura, Ibaraki-ken 319-1195, Japan

e-mail: suyama. kenya@jaea. go. jp T. Yamamoto

Secretariat of Nuclear Regulation Authority, Roppongi-First Bld. 14F, 1-9-9 Roppongi, Minato-ku, Tokyo 106-8450, Japan

M. Suzuki

Central Research Institute of Electric Power Industry, 2-11-1 Iwadokita, Komae-shi, Tokyo 201-8511, Japan

© The Author(s) 2015

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_6

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

Keywords Fission products • Isotopic composition • Post-irradiation examinations

6.1 Introduction

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 amounts of such isotopes should be evaluated by using burn-up calculation codes. For this purpose, a quantitative analytical method of uranium, plutonium, and fission products of spent fuels has been studied [1,2]. However, it is known that there are several barely measurable elements in such important fission products.

To assay the amount of many fission products, radiation measurement is widely used. Cesium-134 and -137 are typical examples. However, this method is not applicable for isotopes that are important from the aspect of reactivity assessment because they are stable isotopes. For such isotopes, there is the possibility of adopting the isotopic dilution method (IDM), which has been used for measurement of actinides and a few fission products such as neodymium.

In the Japan Atomic Energy Agency (JAEA), thermal ionization mass spectrom­etry (TIMS) has been used for IDM to evaluate the burn-up value of the used fuel. TIMS is one of the most reliable instruments to determine the isotopic ratio and the obtained result is considered to be the reference. However, TIMS needs relatively large amounts of the fuel solution sample and a long time is required to obtain the final results after dissolution of the fuel and preparation of the measurement sample.

The most serious problem is that the important fission isotopes for the reactivity assessment belong to the rare earth elements. Because many of these have the same mass number, we need an efficient chemical separation method and high — performance instruments for measuring the isotopic composition that which should have high sensitivity and resolution. For this reason, the fission products important for reactivity assessment are barely measurable and available data for such fission products are scarce.

JAEA has been active in measuring the isotopic composition of the spent nuclear fuel from the 1980s and the obtained data have been archived in the SFCOMPO database [3], which has been supported by the OECD/NEA databank. Based on this past experience, JAEA launched a development program [4] of measurement of the fission products important for reactivity assessment under the auspices of the Japan Nuclear Energy Safety Organization (JNES) in 2008 and successfully finalized the program in 2012.

In this program, a combined method of chromatographic separation of uranium, plutonium, and fission products from irradiated nuclear fuels was developed. Furthermore, by the introduction of high-resolution inductively coupled plasma mass spectrometry (HR-ICP-MS), the IDM has been applied to lanthanide nuclides. The developed method was applied to the measurement of isotopic composition of used BWR 9 x 9 fuel and evaluation of the burn-up calculation code was carried out [5].

After the accidents at Fukushima Dai-ichi Nuclear Power Plants (hereafter referred to as 1F) of Tokyo Electric Power Company (TEPCO) in 2011, we need a confirmed method to assay the composition of the fuel irradiated in 1F to carry out decommissioning of the Fukushima site. For this purpose, JAEA has a further measurement plan of not only BWR but also PWR used fuel to obtain enough experience to measure the isotopic composition of the irradiated nuclear fuels.

This report summarizes the analytical procedure to measure the amount of fission products isotopes developed in JAEA and the future measurement program.

Shizuka Takai and Kouichi Hagino

Abstract A photo-neutron (y, n) reaction with laser Compton scattering y-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 (y, y), (y, n), and (y,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 1018/s effectively reduces the radioactivity of the target 137Cs.

Keywords 137Cs • Giant dipole resonance • Laser Compton scattering • Photo­neutron reaction • Radioactive wastes • Transmutation

1.1 Introduction

One of the major problems of the nuclear fuel cycle is the disposal of high-level radioactive waste that contains long-lived nuclides such as 129I and high-decay heating nuclides such as 137Cs. After the severe accident at the Fukushima Daiichi

S. Takai (*)

Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1115, Japan e-mail: takai. shizuka@jaea. go. jp

K. Hagino

Department of Physics, Tohoku University, Sendai, Miyagi 980-8578, Japan

© The Author(s) 2015

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_1

Nuclear Power Plant, there is also a problem of 137Cs having been concentrated by treatment of contaminated water. Transmuting such nuclides into short-lived or stable nuclides is one possible way to resolve this problem. Neutron capture reactions have been proposed for transmutations of such fission products. However, the neutron capture cross sections differ significantly from nuclide to nuclide, and this transmutation method is not effective for nuclides with small neutron capture cross sections such as 137Cs.

Recently, photo-neutron (y, n) reactions with laser Compton scattering y-rays have been suggested as an alternative method for nuclear transmutations [1, 2]. Figure 1.1 shows a schematic illustration of this transmutation. This transmutation uses y-rays generated by laser photons backscattered off GeV electrons and photo­nuclear reactions via electric giant dipole resonance (GDR) [3], which has a large cross section for most nuclides. The GDR is a collective excitation of a nucleus that decays mainly by the emission of neutrons, and its total cross section is a smooth function of mass number. Therefore, this method is expected to be effective for transmuting fission products regardless of isotopes.

So far, transmutation with laser Compton scattering for some nuclides has been evaluated only in a simple manner. In this chapter, we investigate more quantita­tively the effectiveness of the transmutation with laser Compton scattering, espe­cially for 137Cs.

Cheol Ho Pyeon

Abstract 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 ([1] [2] [3] [4]Np and [5]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.

Keywords 100 MeV protons • ADS • FFAG accelerator • KUCA • Spallation neutrons • Tungsten target

9.1 Introduction

The accelerator-driven system (ADS) has been considered as an innovative system for the nuclear transmutation of minor actinides and long-lived fission products with the use of spallation neutrons obtained from the injection of high-energy protons into a heavy metal target. At the Kyoto University Critical Assembly (KUCA), a series of ADS experiments [1—5] was carried out by coupling with the fixed-field alternating gradient (FFAG) accelerator [6—8], and the spallation neutrons generated by 100 MeV protons from the FFAG accelerator were success­fully injected into uranium — [1, 2, 4] and thorium-loaded [5, 7] cores.

In the ADS facility at KUCA, reactor physics experiments are being carried out to study the neutronic characteristics through the measurements of reactor physics parameters, including reaction rates, neutron spectrum, neutron multiplication, subcriticality, and neutron decay constant. Among these, neutron multiplication was considered as an important index to recognize the number of fission neutrons in the core induced by the external neutron source.

The mockup experiments [5] of thorium-loaded ADS carried out by varying the neutron spectrum and the external neutron source were aimed at investigating the influence of different neutron profiles on thorium capture reactions and the prompt and delayed neutron behaviors in the subcritical system. The results provided important effects of the neutron spectrum and the external neutron source on both static and kinetic parameters: the effect of the neutron spectrum was investigated by varying the moderator material in the fuel region, and that of external neutron source by injecting separately 14 MeV neutrons and 100 MeV protons into the thorium-loaded core varying the moderator. Before the subcritical experiments, a thorium plate irradiation experiment was carried out in the KUCA core to analyze the thorium capture and fission reactions in the critical system as a reference of the subcritical system, although the feasibility of Th capture and U fission reac­tions could be examined in the subcritical state.

In this chapter, experimental results of the uranium — and thorium-loaded ADS are shown. Accuracy was evaluated through the comparison between the experi­ments and the calculations of the Monte Carlo analyses through the MCNPX [9] code with ENDF/B-VII.0 [10], JENDL/HE-2007 [11], and JENDL/D-99 [12] libraries. The ADS static and kinetic experiments at KUCA are presented in Sect. 9.2, the results and discussion of the experiments and calculations in Sect. 9.3, and the conclusion of the study in Sect. 9.4.

Jun-ichi Hori, Tadafumi Sano, Yoshiyuki Takahashi, Hironobu Unesaki, and Ken Nakajima

Abstract 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 pm 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 pm thickness to a gold foil of 10 pm 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.

Keywords Fuel debris • Neutron resonance absorption • Pulsed-neutron source • Self-indication method • TOF

J. Hori (*) • T. Sano • Y. Takahashi • H. Unesaki • K. Nakajima Research Reactor Institute, Kyoto University, 2-1010, Asashironishi, Kumatori, Osaka 590-0494, Japan e-mail: hori@rri. kyoto-u. ac. jp © The Author(s) 2015

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_3

3.1 Introduction

It is surmised that melted fuel debris is present in the cores at units 1, 2, and 3 of Fukushima Daiichi NPP. Identifying the fuel debris status in the reactors is one of the most important issues for decommissioning. Therefore, we need to determine how to analyze the properties of actual debris collected from those cores in advance of removal work. As the debris contains melted fuel and cladding tube and structure materials heterogeneously in addition to salt content, nondestructive assaying of the distribution of nuclear materials within the debris is absolutely essential for nuclear material accountancy and critical safety.

Neutron resonance densitometry (NRD) [1] with the time-of-flight (TOF) tech­nique based on neutron resonance transmission analysis (NRTA) [2] and neutron resonance capture analysis (NRCA) [3, 4] is a promising way to characterize the debris. However, there are two difficulties in applying those methods to fuel debris. In NRD, many resonances of other nuclides that are contained in the debris may make it difficult to identify and quantify the target nuclide. In NRCA, it is expected that the intense decayed gamma rays from debris result in high background and large dead time of the gamma-ray detector. In this work, we propose a new concept of the “self-indication method” as a complementary assay to overcome those difficulties. In the self-indication method, we set an indicator consisting of target nuclide with a high purity at the beam downstream from a sample. By detecting the reaction products such as neutron capture y-rays or fission products from the indicator with the TOF method, the transmission neutron can be measured indi­rectly. The self-indicator is a transmission neutron detector that has high efficiency around the objective neutron resonance energies of the target nuclide, enabling us to quantify effectively the amount of resonance absorption of the target nuclide. Moreover, it is not easily affected by the decayed y-rays from the debris.

In this work, experimental validation for application of the self-indication method was carried out. A part of the preliminary results is shown in this chapter.

Gen Ariyoshi, Daisuke Ito, and Yasushi Saito

Abstract 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.

Keywords Accelerator-driven system • Electromagnetic probe • Four-sensor probe • Lead-bismuth • Turbulence characteristics • Two-phase flow • Void fraction

G. Ariyoshi (*)

Graduate School of Energy Science, Kyoto University, Kyoto, Japan e-mail: ariyoshi. gen.46n@st. kyoto-u. ac. jp

D. Ito • Y. Saito

Kyoto University Research Reactor Institute, Osaka, Japan © The Author(s) 2015

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_11

11.1 Introduction

The accelerator-driven system (ADS) has been developed as the next-generation nuclear energy system and is expected to be used as a nuclear transmutation process

[1] . ADS is a hybrid system that consists of a high-intensity proton accelerator, a nuclear spallation target, and a subcritical core. Lead-bismuth eutectic (LBE) is considered as an option of the spallation target and can also be used as the coolant of the reactor. Neutrons are produced by a nuclear spallation reaction between the protons supplied from the accelerator and the LBE target, and a chain reaction of nuclear fission can then be maintained by the contribution of spallation neutrons. The chain reaction in the core will stop when the supply of protons stops. Therefore, the ADS has a higher safety margin, in principle, than other nuclear energy systems.

As research toward the development of the ADS, a subcritical reactor physics study, a reactor thermal-hydraulics study, and studies on the material, accelerator, and fuel for the ADS have been carried out. However, safety assessment is very important in preparation for a possible severe accident. A pipe rupture in a steam generator is one of the severe accidents of a LBE-cooled ADS. In this case, the direct contact between the LBE and the water ejected from the ruptured pipe of the steam generator might lead to LBE-steam two-phase flow in the reactor pool. If the gas bubble comes into the fuel region, the core reactivity might be affected. Thus, the gas-liquid two-phase flow appearing in the ADS core should be understood in taking measures for such an accident. The gas-liquid two-phase flow in an ADS has density ratio that is an order larger than that of air-water two-phase flow. Although flow models of gas-liquid two-phase flow with a large liquid-to-gas density ratio are required for severe accident analysis, there are fewer studies on two-phase flow in the physical property range of large density ratio mixture. Thus, an experimental database on two-phase flow properties in two-phase flow with a large density ratio should be built and the two-phase flow model should be developed based on the database. In this study, an LBE two-phase flow was measured by using a four­sensor electrical conductivity probe and a miniature electromagnetic probe, and knowledge of the flow structure and the turbulent characteristics in two-phase flow with a large density ratio was obtained.

The objective fission products required for reactivity assessment are samarium, europium and gadolinium. Cesium-133 is also required. Of the important fission products, several metallic isotopes exist in the barely dissolved residue: 97Mo, 99Tc, 101Ru, 103Ru, and 109Ag.

We decided to adopt the isotopic dilution method (IDM) and the calibration curve method to measure the amounts of the stable fission products. For this purpose, we introduced the high-resolution inductively coupled plasma mass spec­trometry (HR-ICP-MS), ELEMENT2 of Thermo Fisher Scientific (Photo 6.1). This instrument has very high sensitivity and enough precision and accuracy to measure the isotopic ratios of objective elements belonging to the rare earth elements.

In this technical development, five samples taken from ZN2 (average burn-up is

35.6 GWd/t) and ZN3 (average burn-up is 53.5GWd/t) fuel assemblies of used fuel of Fukushima Dai-ni Nuclear Power Plant Unit 1 (2F-1) were used for demonstrat­ing the measurement method. Sample positions are shown in Figs. 6.1,6.2, and 6.3. Five fuel samples taken from ZN2 and ZN3 fuel assemblies were dissolved initially in 3 M nitric acid solution at about 110 °C, then the dissolution residue was dissolved again in mixed solutions of nitric, hydrochloric, and sulfuric acid at 180 °C.

Before the measurement of isotopic ratio, the isobar should be separated to avoid contamination. Figure 6.4 shows a schematic of chemical separation. The dissolu­tion solutions of spent fuels were filtrated and the filtrate solution was fed to an anion-exchange resin of UTEVA (Eichrom, USA) to separate U, Pu, and Nd individually. Figure 6.5 shows the yields of lanthanide in each fraction eluted from the Ln resin column in the separation experiment using a simulated dissolution solution of spent fuel. U and Pu in the solution were effectively separated from the solution with more than 95 % efficiency. The eluate solution from the UTEVA resin column was fed to the Ln resin column. Lanthanides elements were separated with hydrochloric acid solutions in the Ln resin column.

Photo 6.1 High-resolution inductively coupled plasma mass spectrometry (HR-ICP-MS) intro­duced in JAEA for the measurement of fission product nuclides [6]

Future Plans

Initially, this measurement technique and procedure have been developed to obtain experimental data for the demonstration of the neutronics calculation code. How­ever, after the accident at the Fukushima Dai-ichi Nuclear Power Plant in 2011, it is recognized that the measurement of the isotopic composition of the used nuclear fuel is crucial to take countermeasures to the accident. We expect that measurement of the amounts of many varieties of isotopes is required for the decommissioning of the Fukushima Dai-ichi site.

To accumulate experience and recheck the measurement procedure, JAEA has already started the measurement campaign after the Fukushima accident. The first PIE sample was taken from the same fuel assembly used in the PIE campaign described in the earlier sections. In 2012, one fuel sample was taken from the ZN3 fuel assembly irradiated in Fukushima Dai-ni nuclear power plant unit 1 and the dissolution was conducted in 2013. It is expected that the measurement results will be obtained in F. Y. 2014.

JAEA will also assay the isotopic composition of spent nuclear fuel irradiated in PWR. It is planned to measure the isotopic compositions of nine samples taken from NO4F69 fuel assembly irradiated in Ohi Nuclear Power Station unit 4 of the Kansai Electric Power Co., Inc. (KEPCO). The nine PIE samples will be taken from three fuel rods including UO2 and UO2-Gd2O3 whose average burn-up values are estimated to be from 40 to 57 GWd/t approximately. This measurement campaign was started in 2013, and the first results are hoped to be seen in 2014.

6.3 Conclusion

The Japan Atomic Energy Agency had been active in the field of postirradiation examinations since the 1980s. Based on past experience and introducing the state — of-the-art technique, JAEA began a measurement program of fission products that are important for reactivity assessment. By this program, the quantitative analytical method based on isotopic dilution technology has been developed for fission products in spent fuels. JAEA will carry out the measurement program of isotopic composition of the used nuclear fuel for countermeasures to the accident at Fukushima Dai-ichi Nuclear Power Station. This program consists of the measure­ment of several samples not only from BWR but also PWR. The obtained results will be used for the evaluation of the burnup code system. Also, the experience of treating spent fuel and measuring its isotopic composition will strengthen the technical ability of JAEA for providing countermeasures for the Fukushima accident.

Acknowledgments The authors thank the following staff in charge of the measurement program in JAEA: K. Tonoike, M. Amaya, M. Umeda, T. Sonoda, K. Watanabe, N. Shinohara, M. Ito,

T. Ueno, M. Magara, J. Inagawa, S. Miyata, S. Sampei, K. Kamohara, M. Sato, H. Usami, K. Ohkubo, M. Totsuka, Y. Sakazume, and T. Kurosawa. The authors are deeply indebted to T. Ichihara and T. Nakai of the Kansai Electric Power Co., Inc., and M. Kawasaki and I. Hyodo of Tokyo Electric Power Co., Inc., who allowed us to use their spent fuel assemblies and to use owned technical data. The authors also like express their appreciation to Y. Taniguchi, H. Nagano, T. Ito, H. Kishita, Y. Kubo, and K. Kakiuchi of Nuclear Fuel Industries, Ltd., for their cooperation in using their spent fuel assemblies and technical data.

Open Access This chapter is distributed under the terms of the Creative Commons Attribution Noncommercial License, which permits any noncommercial use, distribution, and reproduction in any medium, provided the original author(s) and source are credited.

1.2.1 Reaction via Giant Dipole Resonance

Nuclear transmutation with laser Compton scattering uses photonuclear reactions via GDR because the cross section of GDR is quite large and the total cross section is a smooth function of mass number. GDR is a collective excitation of a nucleus involving almost all nucleons, which is interpreted classically as a macroscopic oscillation of a bulk of protons against that of neutrons. The total cross section °gDR, the resonance energy ER, and the width rR are given by [4]

4dR = 60(1 + к) NZ mb • MeV,

Er = 31.2A~1/3 + 20.6A~1/6MeV, (1Л)

Tr = 0.0026ER91 MeV,

where N and Z are the neutron and proton numbers, A = N+Z is the mass number, and к, which is roughly equal to 0.2 for medium nuclides, is a correction coefficient for the pion exchange.

When a target nucleus is irradiated with a photon and excited to GDR, it often forms a compound nucleus with only a small contribution of a pre-equilibrium reaction [5]. The compound nucleus is an excited state in which the energy brought

Fig. 1.1 Schematic illustration of nuclear transmutation with laser Compton scattering

by the incident particle is shared among all degrees of freedom of the nucleus. The reaction cross section from an initial channel a to a final channel в proceeding through a compound nucleus state of spin J can be written by the Hauser-Feshbach formula as

(1.2)

where ka is the wave number in the initial channel, gJ is a statistical factor, T is a transmission coefficient, and (T) is the energy average of T. The statistical factor is

where ia and Ia are the projectile and target spins.

Calculations of reaction cross sections are performed using the nuclear model code TALYS (version 1.4) [6]. The neutron transmission coefficients are calculated via the global optical potential [7]. The gamma-ray transmission coefficients are calculated through the energy-dependent gamma-ray strength function according to Brink [8] and Axel [9]. We employed the level density given by Gilbert and Cameron [10].

Figure 1.2 shows the photonuclear reaction cross sections of 137Cs calculated using the TALYS code. In the incident photon energy B(n) < EY < B(2n), where B (n) and B(2n) are the one — and two-neutron binding energies, respectively, we can see that the (y, n) reaction mainly occurs. Because the resonance energy of GDR ER is 15-18 MeV, which is roughly equal to B(2n) for medium nuclides, about half the reactions via GDR are (y, n) reactions, which occur at B(n) < EY < B(2n).

9.2.1 Uranium-Loaded ADS Experiments

KUCA comprises two solid polyethylene-moderated thermal cores designated A and B and one water-moderated thermal core designated C. The A-core is mainly used for experiments of ADS basic research. The three cores are operated at a low mW power in the normal operating state; the maximum power is 100 W. The constitution and the configuration of the cores can be altered easily, and the coupling with the conventional Cockcroft-Walton type accelerator and with the FFAG accelerator has allowed conducting experiments separately with the use of 14 MeV neutrons from deuteron-tritium fusion reactions and 100 MeV protons with the heavy metal target, respectively.

m Normal fuel (3/8"P36EU)| I | Fe + polyethylene □ Polyethylene reflector H Void x3 + Fe + polyethylene © Control rod И Polyethylene + boron (10wt%)

Safety rod |_b^ Voidx 3 + polyethylene + boron (10wt%)oid

© Neutron source (Am-Be) Ш Voidx 1 + polyethylene (FC Fission chamber | s | Voidx 3 + polyethylene

UIC detector □ Aluminum sheath

I F ‘I Partial fuel (3/8"P14EU) О 1’ty3He detector SV SV Fuel (3/8"P32EU) О 1/2"ф BF3detector

T ungsten target

The A-core (Fig. 9.1) employed in the ADS experiments was essentially a thermal neutron system composed of a highly enriched uranium fuel and the polyethylene moderator/reflector. In the fuel region, a unit cell is composed of the highly enriched uranium fuel plate 1/16 in. thick and polyethylene plates 1/4 in. and 1/8 in. thick (Fig. 9.2). The SV assembly is composed of a

5.8 x 5.08 x 5.08 cm center void region, 32 fuel unit cells, and the polyethylene blocks. In these ADS experiments, three types of fuel rods designated as the normal, partial, and special fuel SV were employed. For reasons of the safety regulations for KUCA, the heavy metal target was located not at the center of the core but outside the critical assembly. As in the previous ADS experiments with 14 MeV neutrons,

the introduction of a neutron guide and a beam duct is requisite to lead the high — energy neutrons generated from the heavy metal target to the center of the core as much as possible. In the uranium-loaded ADS experiments, the proton beam parameters were 100 MeV energy, 0.01 nA intensity, 30 Hz pulsed frequency, 100 ns pulsed width, and 80-mm diameter spot size at the tungsten target (100 mm diameter and 9 mm thick). The level of the neutron yield generated at the target was around 1.0 x 106 1/s by the injection of 100 MeV protons onto the tungsten target.

To verify the self-indication method, we have performed three kinds of experiments using a 46 MeV electron linear accelerator (linac) at the Kyoto University Research Reactor Institute. The experimental arrangement is shown in Fig. 3.1. Pulsed neutrons were produced from a water-cooled photo-neutron target assembly, 5 cm in diameter and 6 cm long, which was composed of 12 sheets of tantalum plates with total thickness of 29 mm [5]. This target was set at the center of an octagonal water tank, 30 cm long and 10 cm thick, as a neutron moderator. The linac was operated with a repetition rate of 50 Hz, a pulse width of 100 ns, a peak current of 5 A, and an electron energy of about 30 MeV. We used a flight path in the direction of 135 ° to the linac electron beam. To reduce the gamma flash generated

by the electron burst from the target, a lead block, 7 cm in diameter and 10 cm long, was set in front of the entrance of the flight tube.

First, a gold foil 10, 20, 30,40, or 50 qm thick was located as a sample at a flight length of 11.0 m, where the neutron beam was well collimated to a diameter of 24 mm. An indicator located at a flight length of 12.7 m was surrounded by a Bi4Ge3O12 (BGO) assembly, which consists of 12 scintillation bricks each 5 x 5 x 7.5 cm3 [6]. Prompt-capture у-rays from the indicator were detected with the BGO assembly in the TOF measurement. A 10B plug 8 mm thick or gold foil 50 qm thick was used as an indicator. Because the former thick indicator can absorb most neutrons with energies below the epi-thermal region, it was equivalent to the conventional NRTA. In the latter case, it was the self-indication measurement. The area densities of the samples with different thickness were estimated by area analysis for the 4.9 eV resonance of 197Au.

As the next step, a 50-qm-thick silver foil was added to a 10-qm-thick gold foil to form a sample and the area density of the gold foil was measured. It is worth noting that silver has a large resonance at 5.2 eV, close to the 4.9 eV resonance of 197Au. The 10B plug 8 mm thick or a gold foil 50 qm thick was also used as an indicator. Moreover, we demonstrated a nondestructive assay for nuclear materials using a mixture composed of a natural uranium plate and sealed minor actinide samples of ‘Np and Am. The natural uranium plate was 10 x 20 mm and weighed 5.8 g. The samples of Np and Am were oxide powder, which was pressed into a pellet 20 mm in diameter and encapsulated in an aluminum disk­shaped container 30 mm in diameter with 0.5-mm-thick walls. The activities of 7Np and Am were 26 and 868 MBq, respectively. In the third measurement, the

10B plug 8 mm thick or a natural uranium plate of 10 x 20 mm2 and weight 5.8 g was used as an indicator.