It is extremely important to shorten the lifetimes of many radioactive nuclei in nuclear wastes. The role of nuclear transmutation technology is one of the main themes of this Symposium. The accelerator-driven system is one of the most promising methods to transmute radioactive nuclei to those of shorter lifetimes.

In Japan, the Omega project, which includes an accelerator-driven system, has been discussed for more than 10 years. I have helped to establish the J-PARC because one of its purposes is to develop the transmutation technology.

I expect that Dr. Hiroyuki Oigawa will tell us about the accelerator-driven system.

I would like to learn about the present situation of the transmutation technology in Japan and in the world.

23.5.3 The Research and Development of Nuclear

Technologies for Reactor Decommissioning, Safety Technology, Back-end, etc., Must Be Promoted Intensively Through International Cooperation

Nuclear technologies for reactor decommissioning, safety technology, back-end, etc., must be urgently developed. They are very important, especially in Japan after the Fukushima Daiichi Accident.

These technologies, however, are also desired in all countries that already have nuclear power stations, and also in countries which are planning nuclear power stations. These technologies therefore should be researched and developed through international cooperation. Fukushima would be a very good candidate for us to construct an international center for researching and developing technologies for reactor decommissioning.

23.2 Conclusion

For the future of human beings, nuclear technology is indispensable to guarantee the safety of energy and to reduce CO2 in the atmosphere, which causes global warming.

For promoting nuclear technology, we must encourage young researchers to be interested in nuclear science and engineering. Education is very important for this purpose.

You who are experts in nuclear science and technology should be very proud of your specialty. It is the most important time for you to solve very difficult problems after the accident of the Fukushima Daiichi Nuclear Power Station. I sincerely hope that you will overcome this crisis caused by the Fukushima accident.

Let us change the misfortune into good luck for the future of human beings.

I hope that this Symposium will succeed in producing good fruits.

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.

The authors proposed a work flow for selecting a waste management technology [2] (as shown in Fig. 28.1). Establishment of the development criteria for each tech­nology is a fundamental issue and the first step of the development.

For treatment technology, it is important to have a simple system that can be applied to a variety of wastes, and which has a volume reduction factor and is economical. In addition, the system must minimize secondary waste and address the difficulty of residual research and development (R&D) to commercialize the technol­ogy. Regarding the stability of waste forms produced by any treatment technology, a low leaching rate is required, especially for waste containing long-lived radionuclides over a certain amount. Easy identification and measurement of the radionuclides in the waste by the treatment process would also be considered an advantage.

Storage technology has many options that are either in operation or under R&D in different countries. The storage cost is an important index to use when selecting an option. If a waste generates a large quantity of heat as a result of containing a large amount of beta — and gamma-emitting nuclides such as Cs-137, the waste storage period should be considered before final disposal to reduce heat generation. If the waste has greater heat generation in disposal, the waste emplacement area must be increased to reduce the temperature of the surrounding engineered barriers or rock to less than the allowable temperature. As a consequence, disposal cost will increase. The storage period and disposal cost have a strong mutual relationship. If the storage period must be increased by more than a few hundred years to reduce heat generation for disposal, certain radionuclides such as Cs-137, whose half-life is less than several decades, would have undergone significant decay by time of disposal, which also raises the possibility of lowering the waste classification level for disposal.

Fig. 28.1 Flow chart for selection of waste management technology [2]

Disposal technology also has many options that are either in operation or under R&D in different countries. The disposal cost depends on the design of the disposal facility, the composition of the engineered barriers, and the depth of the disposal facility, which are strongly dependent on the characteristics of the radionuclides in the waste.

Table 28.2 Items for the management database of treatment technology options I. D. No.

Name of technology Description of technology Applicability to various wastes Process flow diagram Description of process flow diagram

Treatment conditions, including temperature, pressure, conditions, material balance, radionuclide balance, and decontamination factor (DF)

Chemical reagents and utilities Volume reduction ratio of waste

Characteristics of waste form, including configuration, uniaxial compression strength, vacant volume ratio, apparent density, leaching rate of radionuclides, and leaching rate of chemical components Content and status of NaCl

Status of technology; commercialization, development, fundamental research Technical issues (R&D for commercialization)

Standards to be applied References

Technology options for treatment, storage, transportation, and disposal have been proposed by domestic and international organizations. These options should be integrated and managed in a database of technology options. R&D results should also be added in the database as soon as possible. To manage the technology options, a database should be prepared and used to share information among the related organizations. The authors proposed a set of items to be managed in the database of treatment technology options (Table 28.2).

When an appropriate technology is being selected from several options, it is important to evaluate each technology option and compare them. At the end, it is also important to evaluate the combination of technologies from treatment to disposal to make the final selection of an appropriate set of these technologies and to establish a total system for specific wastes. The authors propose a set of indexes for evaluation of the combination of waste management technologies (Table 28.3). Evaluation of the total system should include the long-term safety of disposal because this is the most important issue and goal of waste management.

Table 16.1 and Fig. 16.3 show the core specification and layout with hydride targets. The core layout is based on the Japanese prototype fast reactor Monju. The thermal power is 714 MWt, the diameter of the active core is about 1,800 mm, and the height is 930 mm. The 54 hydride MA-hydride target assemblies are located at the inner most row in the three radial blanket rows. Each assembly contains 61 MA-hydride target pins, where the diameter of the pellets is set at 10.4 mm and the stack length is 930 mm. The ratio of H/M (M=MA+Zr) is considered to be 1.6. The composition of MA is assumed that derived from the typical large LWR discharged fuel, that is, 237Np/241Am/242mAm/243Am/243Cm/244Cm/245Cm/246 Cm = 0.5200/0.2493/0.0010/0.1663/0.0006/0.0592/0.0031/0.0006.

The sensitivity coefficients shown in Sect. 20.3.3 are valid within the assumed analysis conditions in Sect. 20.3.1. However, the impurity elements that are not specified in the standard specification can be possibly present in the material. To know the effect of the difference in the initial composition on sensitivity coeffi­cients, additional analyses were conducted using the initial composition based on measured data. The evaluation of activation products in SUS304 stainless steel is described here.

Table 20.10 Concentration of activation products in SUS304 stainless steel

Except for the initial composition, the analysis conditions described in Sect. 20.3.1 were assumed. The composition data reported by the Atomic Energy Society of Japan [6] were applied in this analysis. In this reference, the concentra­tion distributions of some elements with their mean values and standard deviations have been determined based on several measured data. The initial composition based on measured data is shown in Table 20.9 together with that based on the standard specification.

The concentration of activation products using the initial composition based on measured data is shown in Table 20.10. As a matter of course, the concentrations were changed from those in Table 20.6b because the different initial compositions were assumed. It was found that Nb-94, Tc-99, Mo-93, K-40, and Zr-93 appeared in Table 20.10 because of the presence of niobium, molybdenum, and potassium in the initial composition. For the comparison with the sensitivity coefficients shown in Sect. 20.3.3, sensitivity analyses of cross sections were conducted for the several nuclides Ni-59, Ni-63, Fe-55, Co-60, Mn-54, C-14, and Cl-36, which were listed in both Tables 20.6b and 20.10.

The sensitivity coefficients of cross sections using the initial composition based on measured data are shown in Table 20.11. It was found that the results of Co-60, C-14, and Cl-36 are much different from those in Table 20.8b, which indicates that the dominant generation pathways of these nuclides were changed. Figure 20.3 shows the comparison of dominant generation pathways of Co-60, C-14, and Cl-36 between different analysis conditions. The source nuclides of Co-60, C-14, and Cl-36 were Co-59, N-14, and Cl-35, respectively, under the conditions based on measurement data, whereas those were Ni-60, C-13, and S-34, respectively, under the conditions based on the standard specification.

As shown in the foregoing example, the dominant generation pathway can be changed corresponding to the initial composition. The reliable measured data of initial impurity elements should be used if they are available. For any condition, sensitivity analyses on the basis of the methodology stated in this study can systematically identify the dominant generation pathways of activation products.

Akemi Yoshida

Abstract Although nuclear power has become recognized as a social issue—one that concerns us all—there is still, in Japan, insufficient public debate on the problems posed by this form of energy. In particular, interest among the younger generation on this and many other issues is limited, a situation reflected in the low turnout of young people at elections. The disposal of high-level radioactive waste is an issue that cannot be simply solved by shutting down nuclear reactors. Yet, in spite of the need to urgently find a solution to the problem of nuclear waste, many young people appear to be apathetic. Part of the reason for this lack of interest is that students majoring in the so-called humanities do not feel confident approaching the issue. As a way to raise such students’ interest in the issue of nuclear waste disposal, debating courses were held in the social science departments of two universities located in Aichi Prefecture, Japan. This chapter reports on these courses, discusses the value and effectiveness of debate in raising awareness of social issues, and assesses potential problems with implementing debating in educational contexts.

Keywords Active learning • Classroom debate • Communication training • Funda­mental literacy for members of society • Nuclear power • Radioactive waste

During the early years of nuclear energy R&D in Europe, between 1960 and 1980, the main experimental projects involving Th fuels were related to the HTRs (DRAGON OECD international project in the UK, ATR and THTR reactors in Germany) and also to an irradiation of Th-MOX fuel in the Lingen BWR in Germany. These projects can be seen as scientific successes, but they were not pursued on a commercial basis because of the priority given in Europe to the development of LWRs (except in the UK, where low-temperature gas-cooled reactors were developed), with UO2 as reference fuel, and, for countries having selected the reprocessing cycle strategy, the recycling of the recovered Pu as MOX fuel.

Afterward, several studies were undertaken to examine worldwide interest in Th. In 1997, M. Lung wrote a report entitled “A present review of the thorium fuel cycle” [1] at the request of the European Commission. Then, in the 4th EURATOM Framework Programme, a review of the benefits of the Th cycle as a waste management option was carried out [2].

As a result of these studies, it was recognized that this option presented major advantages in term of actinides management through the “burning” of excess Pu in a non-U matrix (Th oxide), at least for those countries in Europe that considered Pu as a waste and not a source of energy for future utilization in fast reactors. These assessments opened the door to several European irradiation experiments during the 5th EURATOM Framework Programme using Th-MOX, namely in the KWO PWR in Obrigheim (Germany), in the HFR MTR in the Netherland (operated by NRG), and in the BR2 MTR in Mol (SCK^CEN) (“THORIUM CYCLE [3]” and “OMICO

[4] ” projects). These efforts were pursued and completed within the 6th EURATOM Framework Programme, with the demonstration at laboratory scale that this fuel would behave in a comparable way as current MOX fuel (see Sect. 18.3). In the 6th EURATOM Framework Programme, the fuels irradiated in the programs THORIUM CYCLE and OMICO were further investigated (postirradiation examination, radiochemical analysis, and leaching tests) in the “LWR-DEPUTY” project [5] and a strategy study on the “Impact of Partitioning, Transmutation and Waste Reduction Technologies on the Final Nuclear Waste Disposal” (“RED-IMPACT”) was performed [6].

In parallel, efforts at the European level started in early 2000 and are still under way concerning the development of the MSR, using a Th-233U cycle in liquid Th fluoride fuel. Between the 5th and the 7th EURATOM Framework Programmes, several projects (MOST, ALISIA, EVOL) were funded (see Sect. 18.4).

Within the European nuclear research community, a Technology Platform named SNETP (Sustainable Nuclear Energy Technology Platform: www. snetp. eu) gathers most of the stakeholders involved in reactor research. SNETP issued a “Strategic Research Agenda” in May 2009 (revised in 2013 following the Fukushima accident) with an Annex (in January 2011) devoted to Th. In the annex, Th systems are noted as having significant long-term potentialities but also significant challenges before reaching industrial implementation. The two aspects (Pu management, molten salts) mentioned in this chapter were specifically recognized in the Th Annex to the Strategic Research Agenda.

The JAEA research program includes computation of criticality characteristics covering a wide range of fuel debris conditions and validation of the computation by critical experiments. In the former activity, several data sets will be systemat­ically obtained by calculation to establish new criticality safety standards for fuel debris. The new standards will be provided as “criticality maps” that indicate subcritical and critical conditions. The maps also show supercritical conditions that would likely lead to a significant threat of human injury [7]. In the latter activity, the new standards (including computation models) will be validated regarding reactivity worth, coefficients of reactivity, and critical mass by critical experiments with simulated fuel debris samples. A criticality monitoring method­ology will also be studied to improve the criticality control measures for fuel debris.

To pursue the aforementioned critical experiments, the core of the modified STACY has a widely distributed neutron energy spectrum between thermal reactor spectra and intermediate reactor spectra. The neutron energy spectrum of the core can be varied by the lattice pitch of the fuel rods, which range from 10.9 to 25.5 mm, corresponding to a moderator-to-fuel volume ratio ranging from 0.9 to 11. Typical neutron energy spectra of the modified STACY are shown in Fig. 22.2 [8].

This figure also shows typical spectra of hypothetical fuel debris of a BWR fuel pellet (3.7 wt.% 235U, 27.5 GWd/t, 5-year-cooled), for comparison. Both spectra were calculated using a burn-up code, ORIGEN2 [9], and a Monte Carlo code, MVP2

[10] , with a nuclear data library, JENDL-3.3 [11]. It can be seen in Fig. 22.2 that the core spectrum with a lattice pitch of 10.9 mm is equivalent to the debris spectrum in 50 vol.% water. The core spectrum of the modified STACY can cover relatively hard spectra of the fuel debris likely to become critical.

For the measurement of the neutronic characteristics of fuel debris, two sets of experimental equipment should be prepared: one includes reactor material struc­tures simulating fuel debris (zircaloy, stainless steel, concrete, etc.), which are pin-, plate-, or box type and are loaded between fuel rods. The other is a sample-loading device to measure its reactivity and which is installed at a test region in the core tank. The experimental equipment is shown in Fig. 22.3.

Many microorganisms inhabit rice paddy fields, and they are responsible for nutrient cycling. We studied the involvement of microorganisms in environmental transfer of 14C. Microorganisms in batch cultures were treated with autoclaving (121 °C, 15 min), mixing with glutaraldehyde [final concentration of 2.5 % (vol/vol)], and mixing with cycloheximide (final concentration, 250 ^g ml-1). Autoclaving and expose to glutaraldehyde inactivate bacteria and fungi, but expo­sure to cycloheximide only inhibits fungi. The partitioning ratios of 14C into solid, liquid, and gas phases for each treatment sample are listed in Table 26.1. When microorganisms were treated by autoclaving and exposing to glutaraldehyde, almost all the 14C added remained in the liquid phase; that is, negligible transfor­mation of 14C occurred. On the other hand, the 14C atoms in the control and the cycloheximide-treated sample were partitioned into solid, liquid, and gas phases at certain ratios, and these ratios were similar between the control and the cyclohex — imide samples. We confirmed fungi made no contribution to partitioning of 14C

Fig. 26.2 Relationships between pH and partitioning ratios of 14C into the liquid phase (scatter plots). Solid line shows the solubility curve of total carbonic acid in water

Fig. 26.3 Effect of pH on the partitioning of 14C into the liquid phase

based on these results. We concluded that environmental transfer of 14C in rice paddy fields was driven by bacteria, not by fungi.

To confirm incorporation of 14C into bacteria cells, bacteria that were isolated from a flooding water of a paddy soil sample were cultivated on agar plates containing [1,2-14C] sodium acetate [3]. After cultivation, bacterial colonies were formed, and their autoradiography images showed that all colonies had the ability to take up 14C (Fig. 26.4). In our experimental procedure, bacterial cells were conse­quently partitioned into the solid phase, and thus the solid phase contains the 14C incorporated by bacteria, which could be one of the reasons for the relatively high Kd values.

In MOX scenarios, the Rokkasho reprocessing plant (RRP) will be operated with annual capacity of 800 t and a MOX fabrication plant also (Fig. 19.6). The total amount of UO2-SF reprocessed is 34,500 tHM, slightly larger than the planned amount of 32,000 tHM. Thus, the present analysis assumes an extension of the RRP by several years. The MOX loading to a usual LWR is limited to 30 %, although the

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Ohma full-MOX reactor starting in 2014 in this analysis can be operated only by MOX fuel. Because the RRP is operated after all LWRs are closed, part of the separated Pu cannot be burned. The total amount of Pu is reduced to 250 t, but that of MA is increased to 100 t.

Kenji Yamaji

Abstract Concerning the disposal of high-level radioactive waste (HLW) in Japan, the Nuclear Waste Management Organization of Japan (NUMO) has been making efforts toward beginning a literature survey, a first step of HLW disposal according to fundamental policies and final disposal plan based on the “Designated Radioac­tive Waste Final Disposal Act.” However, a difficult situation continues in which responses from municipalities, which are necessary for beginning a literature survey, are not being made.

In September 2010 the Science Council of Japan (SCJ) received a deliberation request from the Chairman of the Japan Atomic Energy Commission, and SCJ formed a Review Committee for Disposal of High-Level Radioactive Waste. The Review Committee made a Reply on Disposal of High-Level Radioactive Waste in September 2012, in which six proposals are made including safe temporal storage and management of the total amount of HLW. In this chapter, an outline of the current HLW disposal policy in Japan and the contents of the Reply are introduced.

Keywords Geological disposal • High-level radioactive waste (HLW) • Risk • Temporal safe storage