The turbulence intensity measured in this study could be divided into wall turbu­lence and bubble-induced turbulence. However, turbulent production from bubbles is dominant at the pipe center. Thus, the turbulence intensity at r/R = 0 was plotted against the void fraction measured by the four-sensor probe (Fig. 11.6). In addition, the present results were compared with the previous experimental data of the bubble-induced turbulence in an air-water two-phase flow system. The solid line in this figure denotes the calculated value by the following semi-theoretical equation [5].

u’ = ur a0’5. (11.2)

In this equation, the velocity field around the bubble is assumed as potential flow and the rotational component of the wake is ignored. In addition, the value calculated by the empirical equation for air-water two-phase flow [6] is also drawn as the dashed line in Fig. 11.6; the equation is represented as follows:

u’ = 0.85a0’8. (11.3)

Although the fluid properties are different with the air-water two-phase flow, the measured turbulence intensity agrees with Eq. (11.3) and the previous data [7—9], except the result at z/D = 3.2. However, Eq. (11.3) was derived for an air-water flow system and its applicability to liquid metal flow was not clear. Therefore, the mechanism of turbulence production in liquid metal two-phase flow should be investigated in more detail. On the other hand, the turbulence intensity at z/ D = 3.2 was slightly larger than other plots and Eq. (11.3). The measurement

Void fraction, a [-]

position at z/D = 3.2 was relatively close to the gas injector, so it is expected that the flow was not fully developed.

11.2 Conclusions

A liquid metal two-phase flow was investigated by using a four-sensor probe and an electromagnetic probe. From the measurement results of two-phase flow structure and turbulence characteristics, the following knowledge was obtained.

• Radial profile of void fraction changes from wall peak to core peak along the flow direction.

• Axial development of the liquid velocity field shows different tendency for the void fraction profiles.

• Existing correlations for interfacial area concentration overestimate interfacial area concentrations at present experimental conditions, which might be attrib­uted to the difference in bubble size. A new correlation should be modeled with further consideration of bubble size and the wall conditions.

• Bubble-induced turbulence at the pipe center in lead-bismuth two-phase flow agrees well with the previous experimental data for air-water flows. However, the mechanism should be clarified by measuring the liquid-metal two-phase flow in a wide range of flow conditions.

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.

Spent nuclear fuel from light water reactors (LWR) contains a mixture of uranium and plutonium (up to 95 % of the initial uranium mass), fission products, and minor actinides such as neptunium, americium, and curium. In the shorter term, the highly active but short-lived fission products will dominate the activity of this spent fuel. However, the transuranics including plutonium and the minor actinides (together with a few long-lived fission products) are largely responsible for the long-term radiotoxicity and heat production of LWR spent fuel.

The principle behind Partitioning and Transmutation (P&T) is to isolate the minor actinides from this LWR spent fuel and transmute them. As for these isotopes the fission to capture ratio increases with increasing neutron energy, a fast neutron spectrum facility is required. By burning the minor actinides, the long-lived, heat — producing component of spent fuel can be strongly reduced, which decreases the radiotoxicity of the spent fuel and its heat load. Both conditions will ease the design and construction of a long-term storage solution (geological disposal) from the engineering point of view.

Partitioning & Transmutation requires the development of an advanced fuel cycle. Currently, two major options for P&T are being studied worldwide: the single-stratum approach wherein the minor actinides are burned in fast reactors that are deployed for electricity production and the double-strata approach where the Pu

is burned for electricity production in LWRs and FRs whereas the minor actinides are burned in a dedicated facility (Fig. 7.5).

In the single-stratum approach, the minor actinides can be mixed homoge­neously in the fast reactor fuel or can be loaded in dedicated targets. In the homogeneous option, care must be taken in the analysis of the change in the core safety parameters such as delayed neutron fraction, Doppler constant, and void coefficient. By increasing the concentration of minor actinides in the fuel mixture, these safety parameters typically go in the wrong direction and hence pose a threat to the reactor safety. Because of this, one expects a maximum of 4-5 % minor actinide loading in the fuel.

Also, the fabrication and reprocessing of this “spiked” fast reactor fuel or the dedicated minor actinide target requires extra care because the presence of the minor actinides increases heat production during these fabrication processes. The presence of Cm-244 will pose a shielding problem because of its spontaneous fission and hence neutron emission.

Given the fact that only small amounts of minor actinides can be loaded per reactor, limited by a maximum concentration in case of the homogeneous option or limited by the number of target positions in the heterogeneous option, a large number of reactors will be required to use this minor actinide-spiked fuel or house these dedicated targets; this will certainly be the case when nations decide to also treat their legacy LWR waste and not only the minor actinides produced in this future advanced fuel cycle. Implied are a large number of transports of these fuels and targets from reprocessing site to fuel fabrication site and to transmutation sites and back.

In the double-strata approach, a dedicated transmutation facility is foreseen in the form of an accelerator-driven system. Because of the reactor physics properties of such an ADS (one does not rely on a subtle equilibrium such as the chain reaction, but the ADS subcritical core acts merely as a multiplier of a primary neutron source), one can devise fuels that have a very high minor actinide content. The EC-FP6 program IP-EUROTRANS delivered the conceptual design of such an industrial transmuter (EFIT). In EFIT, 400 MWth core designs were made with uranium-free inert matrix fuels having a mixture of plutonium and minor actinides. In EFIT, the so-called 42-0 approach core was developed, meaning a core design that would be as plutonium neutral as possible (no burning nor breeding of plutonium) and which could in optimal conditions burn 42 kg minor actinides per TWh power produced. This system was used in the EC-FP6 program PATEROS, which produced a roadmap for the development of Partitioning and Transmutation at the European level. The deployment of such an industrial transmuter as EFIT would be very difficult for small nuclear countries and hence this scheme is optimal in a regional approach.

Because the burning of the minor actinides is done in a very concentrated manner, these industrial transmuters can be located near a fuel reprocessing and transmuter fuel fabrication facility, limiting the transportation of hazardous mate­rials. Calculations have indicated that the support ratio, that is, the ratio of the total power of industrial transmuters to the total power of electricity-generating systems, is about 6 %. Also with this “concentrated” approach, one can much easier envisage the burning of the LWR legacy waste in a reasonable amount of time without impacting the regular electricity production installations.

Within the PATEROS project, a number of nuclear fuel cycle scenarios have been studied. Different regions have been identified: a group of countries that are stagnant with respect to nuclear energy production or in phase-out (“Group A,” typically Belgium, Czech Republic, Germany, Spain, Sweden, Switzerland) and a group of countries which are developing an advanced fuel cycling with the deploy­ment of fast reactors (“Group B,” typically France). Different objectives were set concerning the burning of the minor actinides. Within the EC-F7 ARCAS project, which continues on the work done in PATEROS, it was estimated that to burn the minor actinides present in Group A in a reasonable time frame (less than 100 years), the group would need to deploy 7 EFIT-like facilities. If also Group B wants to stabilize their minor actinide inventory, 15 EFIT-like installations would be needed, and if total minor actinide elimination is required in Groups A and B, 20 EFIT-like installations are to be built.

At the European level, four building block strategies for partitioning and trans­mutation have been identified. Each block poses a serious challenge in research and development to reach an industrial-scale deployment. These blocks are as follows.

• Demonstration of advanced reprocessing of spent nuclear fuel from LWRs,

separating uranium, plutonium, and minor actinides;

• Demonstrate the capability to fabricate at semi-industrial level dedicated transmuter fuel heavily loaded in minor actinides;

• Design and construct one or more dedicated transmuters;

• Demonstration of advanced reprocessing of transmuter fuel together with the fabrication of new transmuter fuel.

MYRRHA will support this roadmap by playing the role of an accelerator-driven system prototype (at reasonable power level) and as a flexible irradiation facility providing fast neutrons for the qualification of materials and fuel for an industrial transmuter. MYRRHA will be capable of irradiating samples of this inert matrix fuels, but it is also foreseen to house fuel pins or even a limited number of fuel assemblies heavily loaded with MAs for irradiation and qualification purposes.

1.2 Conclusions

SCK^CEN is proposing to replace its aging flagship facility, the Material Testing Reactor BR2, by a new flexible irradiation facility, MYRRHA. Considering inter­national and European needs, MYRRHA is conceived as a flexible fast spectrum irradiation facility able to work in both subcritical and critical mode. Despite several nonobvious design challenges, such as the use of LBE, the increased level of seismic loading (consequence of Fukushima), or the choice of passive mode for decay heat removal in emergency conditions, we found no significant showstopper in the design. The R&D program that is running in parallel has taken into account international recommendations from experts concerning the remaining technolog­ical challenges as mentioned in Section VI (above).

MYRRHA is now foreseen to be in full operation by 2025, and it will be able to be operated in both operation modes, subcritical and critical. In subcritical mode, it will demonstrate the ADS technology and the efficient demonstration of MA in subcritical mode. As a fast spectrum irradiation facility, it will address fuel research for innovative reactor systems, material research for GEN IV systems and for fusion reactors, radioisotope production for medical and industrial applications, and industrial applications, such as Si-doping.

The MYRRHA design has now entered into the Front End Engineering Phase, covering the period 2012-2015. The engineering company that handles this phase has currently started the work. At the end of this phase, the purpose is to have

• Progressed in such a way in the design of the facility that the specifications for the different procurement packages of the facility can be written,

• Adequately addressed the remaining outstanding R&D issues,

• obtained the construction and exploitation permits, and

• Formed the international members’ consortium for MYRRHA.

Belgium and SCK^CEN have opened participation in the MYRRHA to EU member states and to the European Commission but also to worldwide participation, as the issue of safe and efficient management of high-level nuclear waste is a worldwide issue, whatever the policy adopted or to be adopted by the countries that have industrialized nuclear power generation and want to phase it out, those willing to continue its use, and those willing to start nuclear power generation.

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.

M. Koizumi, F. Kitatani, H. Tsuchiya, H. Harada, J. Takamine, M. Kureta, H. Iimura, M. Seya, B. Becker, S. Kopecky, W. Mondelaers, and P. Schillebeeckx

Abstract 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 y-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).

Keywords Capture • Fukushima • GELINA • Neutron resonance densitometry • NRD • Nuclear security • Severe accident • Transmission

M. Koizumi (*) • F. Kitatani • H. Tsuchiya • H. Harada • J. Takamine • M. Kureta • H. Iimura Nuclear Science and Engineering Center, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195, Japan e-mail: koizumi. mitsuo@jaea. go. jp

M. Seya

Integrated Support Center for Nuclear Nonproliferation and Nuclear Security, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1118, Japan

B. Becker • S. Kopecky • W. Mondelaers • P. Schillebeeckx

European Commission, Joint Research Centre, Institute for Reference Materials

and Measurements, Retieseweg 111, 2440, Geel, Belgium © The Author(s) 2015

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_2

2.1 Introduction

Quantifying nuclear materials (NM) in the debris of melted fuel (MF) formed in a severe accident is considered to be difficult because of their variety of size, shape, unknown composition, and strong radioactivity. Although techniques of nonde­structive assay (NDA) are indispensable for the evaluation of NM in debris, quantification methods have not been established so far [1]. In the cases of TMI-2 or Chernobyl-4, accounting for the NM was based on some estimations.

We have proposed a technique called neutron resonance densitometry (NRD) [2, 3] to quantify NM in particle-like debris that is assumed to be produced in the rapid cooling processes of a severe accident [4]. Small pieces are also produced when MF are cut or broken down to be taken out of the damaged reactors [1].

To examine the NRD method, studies have begun. Some experiments were carried out at the time-of-flight (TOF) facility GELINA [5] of EC-JRC-IRMM under the agreement between JAEA and EURATOM in the field of nuclear materials safeguards research and development.

In this chapter, we briefly describe the concept of NRD, give an overview of the development of NRD, and explain some parts of the recent progress.

10.2.1 Sample and Irradiation Facility

To investigate surface wettability change by irradiation, samples are irradiated by using an ultraviolet lamp, a 60Co y-ray source, and a proton accelerator. In this study, a TiO2 sample, which is a typical photocatalyst [10], is used to compare the irradiation effects of ultraviolet, y-ray, and proton beam. TiO2 is prepared through anodizing a 0.1-mm-thick titanium plate [11]. Details of the experimental proce­dure with TiO2 samples and irradiation facilities are described as follows.

S. Nakamura, A. Kimura, M. Ohta, T. Fujii, S. Fukutani, K. Furutaka,

S. Goko, H. Harada, K. Hirose, J. Hori, M. Igashira, T. Kamiyama,

T. Katabuchi, T. Kin, K. Kino, F. Kitatani, Y. Kiyanagi, M. Koizumi, M. Mizumoto, M. Oshima, K. Takamiya, Y. Toh, and H. Yamana

Abstract 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.

Keywords Activation method • ANNRI • J-PARC • Long-lived fission products • Minor actinides • Neutron capture cross section • Time-of-flight method

S. Nakamura (*) • A. Kimura • M. Ohta • K. Furutaka • S. Goko • H. Harada • K. Hirose

T. Kin • F. Kitatani • M. Koizumi • Y. Toh

Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun,

Ibaraki 319-1195, Japan e-mail: nakamura. shoji@jaea. go. jp

T. Fujii • S. Fukutani • J. Hori • K. Takamiya • H. Yamana

Research Reactor Institute, Kyoto University, 2-1010 Asashiro Nishi, Kumatori-cho, Sennan-gun, Osaka 590-0494, Japan

M. Igashira • T. Katabuchi • M. Mizumoto

Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, O-okayama, Meguro, Tokyo 152-8550, Japan

T. Kamiyama • K. Kino

Faculty of Engineering, Hokkaido University, Kita 13, Nishi 8, Kita-ku,

Sapporo 060-8628, Japan

Y. Kiyanagi

Faculty of Engineering, Hokkaido University, Kita 13, Nishi 8, Kita-ku,

Sapporo 060-8628, Japan

Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8601, Japan M. Oshima

Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun,

Ibaraki 319-1195, Japan

Japan Chemical Analysis Center, 295-3 Sannou-cho, Inage-ku, Chiba-city,

Chiba 263-0002, 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_5

5.1 Introduction

Associated with the social acceptability of nuclear power reactors, it is desirable to solve the problems of nuclear waste management of the long-lived fission products (LLFPs) and minor actinides (MAs) existing in spent nuclear fuels. A method of nuclear transmutation seems to be one of the solutions to reduce the radiotoxicity of nuclear wastes. The transmutation method makes it possible to reduce both the size of a repository for packages of nuclear wastes and the storage risks for the long term. To evaluate the feasibility of development of the nuclear transmutation method, precise nuclear data of neutron capture cross sections for LLFPs and MAs are indispensable.

This chapter presents joint research activities by JAEA and universities for measurements of the neutron capture cross sections for LLFPs and MAs by activation and neutron time-of-flight (TOF) methods.

Toshihiro Yamamoto

Abstract This chapter discusses the theory of higher-order modes in the Feynman

Y 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 Y function and CPSD for an ADS were recently derived in a paper published by the author. These formulae for the Feynman Y function and CPSD are applied to a subcritical multiplying system with a one-dimensional infinite slab geometry in this chapter. The Feynman

Y 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 Y 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.

Keywords ADS • Feynman-a method • Higher-order mode • Monte Carlo • Neutron noise • Power spectral density

12.1 Introduction

In accelerator-driven systems (ADS), fission chain reactions are driven by spall­ation neutrons emitted from a proton beam target. An ADS is quite different from an ordinary nuclear reactor in that it is always operated at a subcritical state. Thus, the safety requirements for reactivity control can be eased in ADSs. The

T. Yamamoto (*)

Kyoto University, Research Reactor Institute, 2-1010 Asashiro Nishi, Kumatori-cho, Sennan-gun, Osaka 590-0494, Japan e-mail: tyama@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_12

subcriticality of an ADS, however, needs to be continuously monitored to maintain its criticality safety. A reactor noise technique such as the Feynman-a method and the power spectral density method can be a potential candidate for monitoring the subcriticality of ADSs. The noise theory in ADSs is different from the classical reactor noise theory in that multiple neutrons are injected from the proton beam target at a single spallation event and pulsed neutrons are emitted deterministically at a constant period. Many theoretical and experimental studies on the noise theory in ADSs have been performed thus far. The theoretical formula for the Feynman-a method or Rossi-a method in ADSs was studied by, for example, Pazsit et al. [1], Pazsit et al. [2], Kitamura et al. [3], and Muiioz-Cobo et al. [4]. Another technique that uses the auto-power spectral density (APSD) or cross-power spectral density (CPSD) was studied by, for example, Munoz-Cobo et al. [5], Rugama et al. [6], Ballester and Muiioz-Cobo [7], and Degweker and Rana [8]. Sakon et al. recently carried out a series of power spectral analyses in a thermal subcritical reactor system driven by a periodically pulsed 14 MeV neutron source at the Kyoto University Critical Assembly (KUCA) [9].

Both the Feynman-a method and the power spectral density method are intended to measure a prompt neuron time-decay constant a of the fundamental mode because the subcriticality is directly related to the fundamental mode a. The measured results, however, are inevitably contaminated by the higher-order mode components. To obtain an accurate knowledge of the subcriticality, the effect of the higher-order modes needs to be quantified in detail.

Endo et al. [10] derived a theoretical formula of the Feynman Y function that considers the higher order modes. Munoz-Cobo et al. [11] also derived a similar theoretical formula from a different approach. Using these formulae, Yamamoto [12, 13] demonstrated quantitative analyses of the spatial — and energy-higher order modes in Feynman Y functions, respectively. In these two works, the Feynman

Y functions were successfully resolved into spatial — or energy-higher order modes. These discussions, however, involved subcritical multiplying systems driven by a neutron source with Poisson character. They did not account for either a periodi­cally pulsed neutron source or its non-Poisson character. Some previous work that considered the higher-order modes in the noise techniques for ADSs has been published (e. g., [6], [7]). In these previous publications, however, the effects of the higher-order modes have not been quantitatively investigated. Yamamoto [14, 15] presented the formulae of the Feynman Y function and CPSD for ADSs that consider the higher-order mode effects. Yamamoto [15] resolved the Feynman

Y functions and power spectral densities into the mode components. Verification of the formulae was demonstrated by comparing the theoretical predictions with the Monte Carlo simulations of the subcriticality measurement in an ADS.

The purpose of the present chapter is to investigate how the subcriticality would affect Feynman Y function and power spectral density. The subcriticality of an ADS differs from design to design. The smaller the subcriticality, the larger the neutron multiplication that can be gained, which, on the other hand, decreases the margin of criticality safety. The subcriticality undergoes a gradual change as the fuel burn-up proceeds. The Feynman Y function and power spectral density emerge differently as
the subcriticality changes. This chapter shows the dependence of subcriticality measurement on its subcriticality, which will contribute to the design of ADSs and planning of subcriticality measurements in the future.

Toshinobu Sasa

Abstract 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.

Keywords Accelerator-driven system • J-PARC • Transmutation • Transmutation Experimental Facility

T. Sasa (*)

Transmutation Section, J-PARC Center, Japan Atomic Energy Agency, 2-4, Shirakata-Shirane, Tokai-mura, Ibaraki 319-1195, Japan e-mail: sasa. toshinobu@jaea. go. 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_8

8.1 Introduction

After the Fukushima accident caused by the Great East Japan Earthquake, public interest in the management of radioactive wastes and spent nuclear fuels has increased. The Science Council of Japan recommends prioritizing research and developments to reduce the radiological burden of high-level wastes by transmu­tation technology.

The Japan Atomic Energy Agency (JAEA) proceeded with R&D to reduce the radiological hazard of high-level wastes by partitioning and transmutation (P-T) technology [1]. In the framework of the J-PARC project, JAEA also promoted constructing the Transmutation Experimental Facility (TEF) to study minor actinide (MA) transmutation by both fast reactors and accelerator-driven systems

[2] . TEF is located at the end of the LINAC, which is also an important component to be developed for future ADS, and shares the proton beam with other experimen­tal facilities used for material sciences, life sciences, and high-energy nuclear physics.

The TEF (Fig. 8.1) consists of two buildings, the Transmutation Physics Exper­imental Facility (TEF-P) [3] and the ADS Target Test Facility (TEF-T) [4]. Two facilities are connected by the proton beam line with a low-power beam extraction mechanism using a laser beam [5]. TEF-P is a facility with zero-power critical assembly wherein a low-power proton beam is available to study the reactor physics and the controllability of accelerator-driven systems (ADS). It also has availability for measuring the reaction cross sections of MA and structural materials, for example. TEF-T is planned as an irradiation test facility that can accept a maximum 400 MeV-250 kW proton beam to the lead-bismuth (Pb-Bi) spallation target. Using these two facilities, the basic physical properties of a subcritical system and engineering tests of a spallation target are to be studied.

R&Ds for important technologies required to build the facilities are also performed, such as laser charge exchange technique to extract a very low power proton beam for reactor physics experiments, a remote handling method to load MA-bearing fuel into the critical assembly, and a spallation product removal method especially for the polonium. The objectives and construction schedule of the facilities, the latest design concept, and key technologies to construct TEF are under way.

2.2.1 The Concept of NRD

Neutron resonance densitometry is a method of a combination of neutron resonance transmission analysis (NRTA) and neutron resonance capture analysis (NRCA) or prompt gamma-ray analysis (PGA). The fundamental principles of NRTA and NRCA are described by Postma and Schillebeeckx [6].

In NRTA, neutron transmission is measured as a function of neutron energy with a TOF technique. Characteristic neutron transmission dips of Pu and U isotopes are observed in the neutron energy in the range of 1-50 eV [7, 8]. Measurements of these transmission spectra can be carried out with a short-flight path TOF system [9,10].

Although strong y-ray radiation from MF samples does not interfere with NRTA measurements, reduction of neutron flux caused by nuclei with large total cross section (such as H, B, Cl, Fe) makes accurate NM quantification difficult. Nevertheless, the quantities of these contained nuclei could not be determined by NRTA only, because these nuclei do not resonantly interact with neutrons in this energy range. To identify and to quantity the composing isotopes, the NRCA/PGA method is required. Characteristic prompt у rays ware utilized. Table 2.1 shows prompt y-rays emitted from nuclei after neutron capture reaction. Most of these discrete prompt Y-rays have significant intensities. The information obtained by NRCA/PGA enables us to determine the appropriate sample thickness and mea­surement time. This information also supports NRTA analysis.

Fig. 2.1 A rough draft of an NRD facility.

The neutron flight path length for NRTA is 5 m and that for NRCA/PGA is 2 m

Ultraviolet irradiates TiO2 by using a commercial UV lamp. Irradiation intensity is measured by an ultraviolet meter and is controlled by changing the distance between the lamp and the sample. The intensity is varied at a range from 0.01

to 5 mW/cm2. The center wavelength of the ultraviolet from this lamp is 365 nm. Figure 10.1 shows a typical irradiation effect on surface wettability change before and after ultraviolet irradiation.

10.2.1.2 Gamma Rays (y-Rays)

The 60Co Y-ray irradiation facility in the Research Reactor Institute, Kyoto Univer­sity, is utilized for y-ray irradiation. The integrated irradiation dose is estimated by an irradiation time and a distance from the Y-ray source. The Y-ray energy of this facility is about 1 MeV (1.17 and 1.33 MeV) and the maximum dose rate is about 15 kGy/h.

10.2.1.3 Proton Beam

The FFAG (fixed-field alternating gradient) accelerator in the Research Reactor Institute, Kyoto University, is utilized for proton-beam irradiation. The energy of the proton beam is set at about 100 or 150 MeV. The maximum beam current of this facility is about 10 nA.

Although accurate data of neutron capture cross sections are necessary to evaluate reaction rates and burn-up times, there are discrepancies among the reported data for the thermal neutron capture cross sections for LLFPs and MAs. As an example of MA, Fig. 5.1 shows the trend of the thermal neutron capture cross section data for 237Np: the discrepancies are about 10 %. Discrepancies between experimental and evaluated data still remain. As for LLFPs, e. g., 93Zr, Fig. 5.2 shows that there

are discrepancies between ENDF/B-VIL0 and JENDL-4.0 evaluations in the region of the thermal neutron energy. Thus, our concern was focused to remeasure the neutron capture cross sections of those LLFPs and MAs.