One of the problems for hydride used in a fast spectrum core is the thermal spike wherein a large power peak occurs in the fuel pins near the hydride used zones. Figure 16.6 shows the intra-assembly radial power distribution in the outermost assembly of the core, that is, the fuel assembly adjacent to the MA-hydride assembly. If a Zr-hydride assembly without MA was used instead of the MA-hydride assembly, a large power peak appeared at the No. 15 pins adjacent to the hydride assembly (shown by pink line of ZrH16 case in Fig. 16.6). In our proposed case, MA-hydride works as an absorber of thermal neutrons, and thus a thermal spike is suppressed. As a result, the radial power distribution of the core has an ordinary profile in the core zones (Fig. 16.7). The power of the first row in the blanket region is, however, a little larger than that of ordinary fast reactors because of the fission reactions of MAs or daughter nuclides, although this power increase is considered to be controllable by adjusting the assembly flow distribution.

Figure 16.8 shows the mass balance of MA for the system of about three 1GWe — class LWRs and one FBR with MA-hydride target as previously described. The LWR annually produces spent fuel with burn-up of 45 GWd/t containing 23 kg MA.. The mass of transmuted MA per year is almost equivalent to that produced annually in about three LWRs, which means that most of the produced MA is

Fig. 16.6 Intra-assembly power distribution of the assembly adjacent to hydride assemblies

transmuted in the system and only a small amount of MA is transferred to the waste stream. As seen in Fig. 16.8, MA recycling is necessary for higher transmutation efficiency. It takes ten times the effective half-life to reduce the mass of MA to 1/1,000 of the initial mass. When the irradiation time of one cycle is 2.19 year, ten fuel cycles are necessary to reduce the mass of MA to 1/1,000.

Kotaro Tonoike, Hiroki Sono, Miki Umeda, Yuichi Yamane, Teruhiko Kugo, and Kenya Suyama

Abstract In the Three Mile Island Unit 2 reactor accident, a large amount of fuel debris was formed whose criticality condition is unknown, except the possible highest 235U/U enrichment. The fuel debris had to be cooled and shielded by water in which the minimum critical mass is much smaller than the total mass of fuel debris. To overcome this uncertain situation, the coolant water was borated with sufficient concentration to secure the subcritical condition. The situation is more severe in the damaged reactors of Fukushima Daiichi Nuclear Power Station, where the coolant water flow is practically “once through.” Boron must be end­lessly added to the water to secure the subcritical condition of the fuel debris, which is not feasible. The water is not borated relying on the circumstantial evidence that the xenon gas monitoring in the containment vessels does not show a sign of criticality. The criticality condition of fuel debris may worsen with the gradual drop of its temperature, or the change of its geometry by aftershocks or the retrieval work, that may lead to criticality. To avoid criticality and its severe consequences, a certain principle of criticality control must be established. There may be options, such as prevention of criticality by coolant water boration or neutronic monitoring, prevention of the severe consequences by intervention measures against criticality, etc. Every option has merits and demerits that must be adequately evaluated toward selection of the best principle.

Keywords Criticality control • Fuel debris • Fukushima Daiichi

21.1 Introduction

In normal nuclear facilities, the goal of criticality control is to secure subcritical conditions of fissile materials, which is achieved by regulating the composition, geometry, or mass of the fissile materials [1]. In the accident of Three Mile Island

K. Tonoike (*) • H. Sono • M. Umeda • Y. Yamane • T. Kugo • K. Suyama Japan Atomic Energy Agency, 2-4 Shirakata Shirane, Tokai, Ibaraki, 319-1195, Japan e-mail: tonoike. kotaro@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_21

Unit 2 reactor (TMI-2), heavily damaged and melted fuel assemblies formed a large amount of fuel debris whose composition was unknown except the possible highest 235U/U enrichment, 3 wt%, whose geometry is uncertain, and whose mass is larger than the minimum critical mass derived from the enrichment. Moreover, the fuel debris had to be cooled and shielded by water. To overcome this uncertain situation, the coolant water was borated with a concentration, >4,350 ppm, sufficient to secure the subcritical condition [2].

The situation of the damaged reactors in Fukushima Daiichi Nuclear Power Station (1FNPS) is more severe than that of TMI-2 because of the water issue. The most major difference is that the coolant water flow is practically “once through.” Boron should be ceaselessly added in the water to maintain its lowest concentration necessary to secure the subcritical condition, which is not feasible. The water is not borated relying on the circumstantial evidence that the xenon gas monitoring in the containment vessels (CVs) does not show a sign of criticality. Although the fuel debris will not be touched for a while, its condition may change because of a gradual drop of its temperature or change of its geometry by aftershocks. The condition will be intentionally changed when the fuel debris is retrieved. Every such change may lead to the criticality of fuel debris [3].

To avoid criticality and its severe consequences, a certain principle of criticality control must be established. There may be options, such as prevention of criticality by coolant water boration or by neutronic monitoring, prevention of the severe consequences of criticality, etc. Each has merits and demerits.

It is necessary to understand the actual condition of the fuel debris regarding the selection of an appropriate principle from those options and the realization of certain criticality control following the selected principle. Adequate observation, sample taking, and analysis of the fuel debris must be conducted.

One reason is that expressing one’s opinion is not necessarily viewed in a favorable light at school and in the home. The way of thinking in Japan, encapsulated in the saying wa wo motte totoshi (harmony is of utmost importance), leads to virtue being placed on conforming to the views of others rather than asserting one’s own opinions. It is partly because of this feeling that Japanese people in general have had relatively little practice in expressing their ideas and opinions to others. In recent years, debate-based lessons have been introduced by some elementary schools, but the spread of such lessons through elementary school education as a whole has not been sufficient. The second reason is rooted in the pivotal role given at the elementary level to sakubun, or essay writing. Pupils practice exploring their emotions and putting them down on paper, but not how to think in a logical way and support their opinion with evidence. A third reason given for the situation described here is the complexity of social problems nowadays, making understanding difficult for “ordinary people,” who just give up even thinking about the issue.

The more complex issues become, however, the greater the necessity for citizens to engage in discussion, express their opinions, and then make decisions. It is, therefore, incumbent on each citizen to grapple with and discuss such issues. Judging from the present situation, it does not seem that staging public debates, having experts offer explanations, or other simple methods can serve as a substitute for real engagement by the citizens. Although workshops aimed at citizens have been held, it is doubtful that participation in debates that occur in such forums is built on a sufficient understanding of the issue at hand. There also appear to be cases when debate is based only on information that fits the administration’s agenda. The information is simplified, and any exchange of views is debate in name only, without substance. It is necessary to look for effective ways to discuss issues to make such workshops and public debates productive and to change the explanatory meetings from superficial gesturing into something more substantive.

Moreover, in spite of the fact that many of the social problems we face are not just a concern for now, but also for the future, many of those interested in social issues and who vote in elections are elderly. Listening to the voices of the elderly is, of course, important, but, for a healthy democracy, it is necessary for young people to engage in debate and be involved in making decisions that affect their society. But how can such engagement by young people be promoted? The way that the author would like to suggest in this chapter is through debating lessons in school.

In a narrow sense, debate can be defined as follows: “discussion on a specific issue involving two groups of speakers, with one group taking a position of supporting the topic and the other arguing against it. Each of the groups seeks to persuade a third party.” [1] (e. g., Yomiuri Shimbun 2013:2). Many of the topics for discussion are chosen from policy issues. To encourage participants to approach an issue from new and multiple perspectives, they are allocated (usually by the teacher) to either one of the groups, that is, they do not choose for themselves which side of the issue to support. This chapter describes a case study of a debating course, taught by the author, and examines its effects and any issues that emerged. Aimed at undergraduate students in a university, the topic of debate in the course was “the problem of high-level atomic waste disposal.”

The MSR, which incorporates the reprocessing on line and needs no specific Th fabrication, adds the benefits of Th without its main challenges. In particular, breeding may be achieved over a wide range of neutron energies, which is not the case for the U-Pu cycle.

Under the European Framework Programs, conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts open promising possibilities to exploit the 232Th-233U cycle. In addition, they can also contribute to significantly diminishing the radiotoxic inventory from present reactor spent fuels, in particular by lowering the masses of transuranic elements. Finally, if required because of expansion of nuclear electricity generation breeding beyond the iso-generation could be achieved. With the Th-U cycle, doubling times values are only slightly higher than those predicted for solid-fuel fast reactors working in the U/Pu cycle (in the range 40-60 years). The characteristics of different launching modes of the MSFR with a thorium fuel cycle have been studied, in terms of the safety, proliferation, breeding, and deployment capacities of these reactor configurations [10].

Between Framework Programmes 5 and 7, several projects (“MOST”, “ALISIA”, “EVOL”) were conducted, and promising developments and results were obtained in particular in the following areas:

• Conceptual design studies

• Safety developments, in particular, to study the residual heat extraction; tests with liquid salts have been undertaken to prove the ability of the cold plug system to act as a security valve on the loop circuit

• Fabrication of the salt mixture (LiF-NaF-KF) to be used in the French molten salt loop (FFFER project) has been achieved

• Experimental investigation of physicochemical properties of fluoride salts

• Experimental tests of the metallic-phase extraction process;

• Corrosion studies and experiments (this remains one of the main challenges for the development of the reactor system)

Finally, it should be noted that the MSR with its Th cycle is one of the six reference systems selected for R&D collaboration in the framework of the Gener­ation IV International forum. The main contributors are the European partners, supported by Russia as observer.

18.2 Conclusions

Since the early 1970s, studies and experimental projects have been undertaken in Europe to examine the potential of Th-based fuels in a variety of reactor types. These projects have all been successful from a scientific point of view, but not all were followed up relative to the overall development of nuclear industry in Europe. High-temperature reactors (HTRs), although very well suited for Th use, have not been deployed to the benefit of LWRs. Results on the use of Th matrices in Th-MOX fuels in LWRs are encouraging, but still need demonstration at a larger scale in commercial conditions. Finally, the probably most efficient use of Th would be in a salt, to feed MSRs. Conceptual studies and related experimental programs are under way.

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 license application for the STACY modification was sent in February 2011 and has been under safety review by the Nuclear Regulation Authority (NRA) of Japan to comply with new safety standards for research reactors enforced in December 2013 [13]. In particular, the NRA will strictly demand prevention measures against natural disasters such as a tsunami from all reactors located at a low altitude. The modified STACY, the reactivity of which is controlled by water level, has a risk of criticality accidents for the duration of tsunami attacks. The prevention measures

Fig. 22.4 Schedule of the STACY modification. CV containment vessel

against criticality accidents are important requirements for the modified STACY: for example, limitation of the core configuration together with the safety plates inserted so as to keep a subcritical state during submersion.

A schedule of the STACY modification is shown in Fig. 22.4. The first criticality experiment in the modified STACY is scheduled for 2018. The modified STACY will provide benchmark data on criticality safety for fuel debris to validate the criticality control measures applicable to the Fukushima Daiichi NPS. The new criticality control measures need to be established by the time the fuel debris begins to be retrieved from each reactor unit of the Fukushima Daiichi NPS. According to the governmental council, retrieval of the fuel debris is scheduled to start as early as 2020, depending on the progress in the decommissioning of each reactor unit [2].

From the aforementioned results, the behavior of 14C in rice paddy fields could be considered as follows (a conceptual diagram appears in Fig. 26.5). When irrigation water is contaminated by 14C-bearing sodium acetate, the 14C compound is taken up and metabolized by indigenous bacteria. A part of the 14C is assimilated by the bacterial cells, and the rest of the 14C is released as gaseous compounds from the cells as a result of dissimilation. The dominant chemical species of 14C in gas forms is carbon dioxide, and thus some of the released 14CO2 is dissolved in soil solution depending on pH. For example, when the pH of the soil solution is less than 6.5, most of 14C in gas forms is released into the air. The released 14CO2 is eventually taken up by rice plants during photosynthesis. When the pH of the soil solution is between 6.5 and 10.5,14C-bearing bicarbonate ion dominates in the soil solution. In addition, once 14CO2 has been released into the air, a part of the 14CO2 gas may be redissolved in the soil solution again as bicarbonate ion. When the pH of the soil solution is greater than 10.5, although this is not probable in paddy fields, 14C — bearing carbonate ion dominates in the soil solution. Carbonate ion is thermally unstable and thus precipitates as carbonate minerals such as CaCO3. In these alkaline situations, the ratio of 14C in the solid phase may increase as a result of the precipitation of 14C. Because the root uptake of 14C by rice plants is negligible, gasification of 14C is an important environmental transfer pathway for the safety assessment of TRU wastes, and bacteria are responsible for driving this pathway.

Acknowledgments This work has been partially supported by the Agency for Natural Resources and Energy, the Ministry of Economy, Trade and Industry (METI), Japan.

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.

This chapter presents issues and measures against the uranium-free TRU metallic fast reactor core. Also, the targets and constraints in parametric survey and selection of core and fuel specification are briefly described.

There are two main issues associated with the TRU burning fast reactor cycle using uranium-free metallic fuel in terms of practicability:

(1) Decrease in the absolute value of the negative Doppler reactivity coefficient resulting from absence of uranium-238, which has the ability to absorb neutrons at elevated temperatures. example,

metallic fuel with uranium: —1 x 10—3 Tdk/dT metallic fuel without uranium: —6 x 10—4 Tdk/dT

(2) Increase in burn-up reactivity swing as fissile decreases monotonically in uranium-free core. example,

metallic fuel with uranium: ~1 %dk/kk’/150 days metallic fuel without uranium: ~6 %dk/kk’/150 days

To solve these issues, there are several candidates, as follows:

(1) Enhance Doppler feedback

— Introduce diluent material in the metallic fuel

— Introduce spectrum moderator

(2) Reduce burn-up reactivity swing

— Reduce the core height

— Introduce neutron absorber outside the core

— Increase the number of refueling batches

Generally, if it is conventional fast reactors with U-Pu fuel, the burn-up reac­tivity swing depends mainly on decrease of fissile amount and increase of neutron parasitic capture of fission products and actinides from burn-up. Therefore, the typical ways to reduce burn-up reactivity swing are to increase conversion ratio via fissile enrichment reduction and to reduce neutron parasitic capture. Here, the conversion ratio is defined as the amount of fissile materials production divided by the amount of neutron absorption, that is, fission and capture, and natural decay of fissile materials. It is difficult, however, for a uranium-free core to increase the conversion ratio because fissile enrichment cannot be controlled in the absence of uranium. Although the reduction of neutron parasitic capture by neutron spectrum hardening improves burn-up reactivity swing, it also harms the Doppler effect. For these reasons, when it comes to uranium-free core, increase of the fissile amount at the beginning of the cycle makes sense because it reduces the ratio of the fissile consumption to the fissile amount at the beginning of the cycle.

These candidates were parametrically surveyed to evaluate the feasibility of the uranium-free TRU metallic fuel fast reactor core in light of aforementioned issues. The targets assumed were the core performances with the Doppler reactivity coefficient equivalent to a conventional U-Pu metallic fuel core. Furthermore, constrains associated with fuel fabrication such as melting temperature was taken into consideration because, in this evaluation, diluent material was assumed to be used as a fuel slug alloy, not cladding material. Hence, the slug was assumed be

Fig. 15.2 RZ geometry for parametric survey fabricated by injection casting as the same as the conventional metallic fuel. This step makes the allowable maximum melting temperature of the fuel alloy less than 1,200 °C to prevent Am volatilization during injection casting [12].

In the ADS scenario, transmuter is changed from FR to ADS. ADS can accept both Pu and MA; distribution is shown in Fig. 19.8. In 2050, 22 ADSs are to be introduced, corresponding to 140 t available TRU. Then, 7 and 3 ADSs are operated respectively from 2110 to 2170 and 2170 to 2230. After three genera­tions, Pu and MA are reduced to 10 t and 3 t, respectively, excluding 16 t MA in vitrified waste.

MA inventory (t)	Pu inventory (t)
—* —* Ю Ю W W 4S. О Ol О Ol О Ol о о о о о о о о
1970 2000 2030 2060 2090 2120 2150 2180 2210 2240 2270
2300
2300
2300

Electoricity generation capacity (GWe)

1970 2000 2030 2060 2090 2120 > b 2150 2180 2210 2240 2270

OOIOOIOOIOOIOOIO

2300

HLW contains very toxic fission products. Fission products in the spent nuclear fuels are highly radioactive. Some countries such as Finland, Sweden, and USA directly dispose spent nuclear fuels as HLW after cooling at spent fuel storage. According to the conventional nuclear fuel cycle policy, spent nuclear fuels in Japan are reprocessed for separating fission products from uranium and plutonium, and the separated fission products are vitrified and then contained in canisters made of stainless steel. The option of direct disposal of spent nuclear fuels was seriously discussed in the first time in Japan at the process for formulating the 2005 Frame­work for Nuclear Energy Policy, and after the Fukushima accident, direct disposal of the spent fuel is becoming a more realistic option.

Right now, 1,984 HLW canisters (vitrified wastes) are stored in Japan. Among the 1,984, 1,442 HLW canisters were sent back from France and UK according to the contracts for the reprocessing commissioned to these countries; the rest are the HLW canisters produced by domestic reprocessing (295 from the test operation of the Rokkasho reprocessing plant and 247 from the Tokai pilot reprocessing plant). An additional 770 HLW canisters will be sent back from the UK, and high-level liquid waste, which is equivalent to 630 HLW canisters, is stored at the Tokai pilot plant.

In addition to the HLW canisters produced by reprocessing, about 17,000 t of spent nuclear fuels is stored at nuclear power plants (about 14,0001 in total) and the Rokkasho reprocessing plant (around 3,000 t). If all these spent fuels are reprocessed at the Rokkasho reprocessing plant, about 21,250 HLW canisters would be added. Thus, even if Japan decided to no longer operate nuclear reactors, we still must dispose HLW equivalent to 24,634 HLW canisters. We cannot run away from HLW issues.

Yoko Fujikawa, Hiroaki Ozaki, Hiroshi Tsuno, Pengfei Wei,

Aiichiro Fujinaga, Ryouhei Takanami, Shogo Taniguchi, Shojiro Kimura, Rabindra Raj Giri, and Paul Lewtas

Abstract Municipal solid wastes (MSW) with elevated concentrations of radioac­tive cesium (rad-Cs hereafter) have been generated in some areas of Japan in the aftermath of the Fukushima Daiichi Nuclear Power Plant (F1 hereafter) accident. Both recycling and final disposal of the contaminated MSW have become a difficult problem in the affected areas, resulting in accumulation of treated residues in the treatment facilities.

The rad-Cs in MSW, especially fly ash, often showed a high leaching rate. Extraction of contaminated MSW with water or hot oxalic acid followed by selective removal of rad-Cs from the extract using ferrocyanide (Fer hereafter) coprecipitation technique could be an ultimate solution for waste volume reduction. The MSW extracts contain various metal components as well as chelating reagents like oxalic acid, and are often very saline. The composition of the extract varies widely depending on waste sources, applied treatment techniques, and rad-Cs extraction method etc. The applicability of the Fer coprecipitation technique had to be tested and validated before it could be applied for actual treatment.

In this work, we applied the Fer technique and observed removal of cesium (Cs) from water and oxalic acid extracts (all spiked with rad-Cs tracer or stable Cs) of various MSW samples collected from uncontaminated areas. Finally, the Fer technique was applied on site for removal of rad-Cs in the extracts of

Y. Fujikawa (*)

Kyoto University Research Reactor Institute, Asahiro-nishi, Kumatori-cho, Sennan-gun,

Osaka 590-0494, Japan

e-mail: fujikawa@rri. kyoto-u. ac. jp

H. Ozaki • H. Tsuno • P. Wei • A. Fujinaga • R. Takanami • S. Taniguchi • R. R. Giri Osaka Sangyo University, 3-1-1 Nakagaito, Daito-shi, Osaka 594-8530, Japan

S. Kimura

Osaka University of Pharmaceutical Sciences, 4-20-1 Nasahara, Takatsuki,

Osaka 569-1094, Japan

P. Lewtas

Edith Cowan University, 270 Joondalup Drive, Joondalup WA6027, Australia © The Author(s) 2015

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_29

contaminated MSW. By modifying coprecipitation conditions according to solution matrix, Cs removal rates of higher than 95 % could be obtained.

Keywords Cesium • Ferrocyanide • Metal • Municipal solid waste • Oxalic acid • pH