1. Professor Daisuke Fujikawa, with the support of NUMO (as was had in this present study), conceived this theme for a debate held in 2012 for undergraduate students on a teacher-training course at Chiba University [3] (Fujikawa, D.2013:5).

2. Sugiyama Jogakuen University, the university to which the author is affiliated, is a participant in a project supported financially by the Ministry of Education, Culture, Sports, Science and Technology. Called the “Project of Educational Reform and Structural Improvement to Respond to the Needs of Industry,” it is composed of 23 universities (including junior colleges) in the central region of Japan. As part of the project, each university engages in career education, and Sugiyama Jogakuen University for its part has focused on active learning to raise its educational performance. The “Fundamental Literacy for Members of Society Questionnaire” was implemented as part of this effort to effectively utilize active learning. The questionnaire consisted of the following 17 items, together with a short definition.

1. Independent (autonomous) learning: The capacity to engage in an activity on one’s own volition.

2. Initiation ability: Appealing and encouraging others to become involved.

3. Seeing-things-through ability: The capacity to set a target and act to achieve it.

4. Topic-finding ability: The capacity to analyze the situation and clarify aims and issues.

5. Planning ability: The capacity to clarify the process that can lead to a solution; the ability to plan.

6. Imaginative ability: The capacity to create new value.

7. Expressive ability: The capacity to convey your views clearly in an easily understandable way.

8. Listening ability: The capacity to listen attentively to what the other person is saying.

9. Flexibility: The capacity to understand other people’s opinions and positions.

10. Grasping-the-situation ability: The capacity to understand the relationship between you and the people and the situation around you.

11. Discipline: The capacity to follow society’s rules and keep promises made with others.

12. Stress-control ability: the capacity to respond appropriately to sources of stress.

13. Sensitivity: The capacity to respond to stimulus from the external environment.

14. A broad education: The possession not just of knowledge, but also an ability to understand and process knowledge creatively.

15. Specialist knowledge and skills: The possession of in-depth knowledge and skills in a particular academic or other field.

16. Logical thinking ability: the capacity to think coherently and logically.

17. Critical thinking ability: the capacity to analyze and judge the suitability and validity of issues and arguments.

Sugiyama Jogakuen University [4] (Sugiyama Jogakuen University 2013) Sugiyama Jogakuen University et al. [5] (Sugiyama Jogakuen University, Special Committee on Career Education, Whole-Faculty Faculty Development Committee 2013)

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.

Two cases of Pu feed were assumed in the present design for Pu transmutation: Pu-ADS, to which only Pu is provided from reprocessing process of LWR SF, and Pu+U-ADS, to which Pu accompanied by U with 50 % weight ratio is provided from the process. Treatment of pure Pu raises proliferation concern in a country without nuclear weapons, and the reprocessing plant is designed to add depleted

Fig. 19.2 R-Z model of accelerator-driven system (ADS)

238 239

U to Pu just after separation. Addition of U results in 9Pu production and

generally is undesirable for Pu transmutation. Table 19.3 lists two compositions. Other conditions are similar to the MA-ADS.

19.3.2 Result of One-Batch Core

The reference ADS for MA transmutation is designed with a one-batch core, which means that all the fuel is loaded and unloaded simultaneously. At first, this one-batch design was adopted to Pu transmuters. Table 19.4 lists the in-core and out-core time with operation efficiency. Out-core time of the ADSs is 3 years, which allows decay of 244Cm. In-core time of the reference MA-ADS is 2 years, although that of Pu-ADS is reduced to 1 year because the decrease of criticality is

too rapid for this ADS. The operation efficiency, eo, is 82.1 % assuming 300 days operation annually.

Design and transmutation performance are summarized in Table 19.5. Volume fraction of the inert matrix, ZrN, of the MA-ADS core is 69.8 %, adjusted so that k-effective at the beginning of the cycle (BOC) of the equilibrium core becomes 0.97. The equilibrium core is obtained after calculating ten cycles of burning, cooling, and recycling. Volume fraction of the Pu-ADS is more and that of the Pu+U-ADS is almost the same. The inventory at BOC of the heavy metal in Table 19.5 is proportional to a one-volume fraction of ZrN. An interesting observa­tion is that the amounts of Pu at BOC are equal among three ADSs, which means U and MA contribute very little to the criticality before depletion. However, impacts on the criticality drop after depletion is significant (Fig. 19.3). kgff drop of the MA-ADS is as small as 1.5 %dk, although others lose 14 %dk even at the equilibrium cycle around 6,000 days, which means MA is a better fertile than 238U. The Pu-ADS has a steeper decrease than Pu+U-ADS because of the absence of 238U. The huge drop of the Pu — and Pu+U-ADS is not acceptable in the current design of accelerator and target for the MA-ADS; the acceptable drop is about 3 %dk in the MA-ADS.

The effective transmutation rate and transmutation half-life are listed at the bottom of Table 19.5. The half-life of the Pu-ADS is shortest because its specific heat is twofold larger than others although its cycle efficiency, ec, is much smaller than others.

It is now very clear that it is almost impossible for renewable energy to replace fossil fuel in the near future. Both nuclear energy and renewable energy are necessary, not only in Japan but also in the world. At the same time we must develop a new technology to compensate for CO2 emissions from fossil fuels.

23.1 Nuclear Technology Must be Developed

23.5.1 Safety Technology of Nuclear Energy Must Be Developed for the Future

Concerning nuclear energy, we must not stop researching and developing new advanced reactors in which greater safety is guaranteed against natural calamities as well as manmade disaster. Small-scale nuclear reactors also should be developed to decentralize electric power stations. If economical problems are overcome, smaller-scale reactors might be easier to guarantee safety.

23.5.2 Technology for the Back-end of the Nuclear Fuel

Hiromi Tanabe and Kuniyoshi Hoshino

Abstract The earthquake and tsunami on March 11,2011, caused severe accidents at the several Fukushima Daiichi Nuclear Power Units, and a significant volume of highly contaminated water was generated from the accident. Several methods have been applied to decontaminate the water, including systems from AREVA S. A. and Kurion, Inc., in addition to the SARRY (Simplified Active Water Retrieval and Recovery System) and ALPS [Advanced Liquid Processing System; incorporated in the MRRS (Multi Radionuclide Removal System)] systems from Toshiba Cor­poration. After the decontamination treatments using these systems, various kinds of sludge and spent adsorbents were generated as secondary wastes. These wastes are now tentatively stored at the site, but further treatment shall be applied to produce appropriate waste forms for interim storage and final disposal in a repository.

Waste management—the treatment, storage, transportation, and disposal of these wastes—is believed to require several decades. The authors examined how to manage these wastes in consideration of the large volume of waste, the variety of waste types, and the long period required to carry out their treatment and disposal in a safe and efficient manner. The requirements for an inventory list and online waste management system; a development strategy for waste treatment, storage, trans­port, and disposal; formation of an R&D implementation and evaluation team; and long-term knowledge management are discussed in this chapter.

Keywords Contaminated water • Disposal • Fukushima Daiichi Nuclear Power Units • Inventory • Secondary waste • Treatment

H. Tanabe (*) • K. Hoshino

Radioactive Waste Management Funding and Research Center, 1-15-7, Tsukishima, Chuo-ku, Tokyo, Japan e-mail: tanabe. hiromi@rwmc. or. 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_28

28.1 Introduction

A significant volume of highly contaminated water was generated from the accidents at the Fukushima Daiichi Nuclear Power Units. Several methods have been applied to decontaminate the water. After decontamination treatments using several systems developed by AREVA, Kurion, and Toshiba, various kinds of sludge and spent adsorbents were generated as secondary wastes. These wastes are now tentatively stored at the site, but further treatment shall be applied to produce appropriate waste forms for interim storage and final disposal in a repository. Management of these wastes is believed to take several decades. The authors examined how to manage these wastes in consideration of the large volume of waste, the variety of waste types, and the long period to carry out the treatment and disposal in a safe and efficient manner. The issues identified are discussed in the following sections.

Kenji Konashi and Tsugio Yokoyama

Abstract A hydride target including minor actinides (MA) is able to enhance the transmutation rate in a fast breeding reactor (FBR) without degradation of core safety characters. Fast neutrons generated in the core region of the FBR are moderated in the MA-hydride target assemblies and then efficiently absorbed by MA. The MA-hydride target pin has been designed in the light of recent research of hydride materials. This chapter shows the feasibility of MA transmutation by an existing reactor, Monju.

Keywords Fast breeder reactor • Hydride • Minor actinide

16.1 Introduction

High-level wastes generated after reprocessing spent nuclear fuels include long — lived radioactive nuclides of minor actinides (MA), such as 237Np, 241Am, 243Am, and 244Cm. A currently available method for the final disposal of the high-level wastes is to vitrify them under rigid control, to store them in monitored locations until the radiation decays to allowable levels, and then to dispose them underground.

The transmutation of MA by the fast breeder reactor (FBR) has been intensively studied to reduce radioactivity of the wastes [1]. Transmutation rate, which is one of the most important factors for transmutation methods, is determined by values of neutron flux and nuclear reaction cross section as follows:

K. Konashi (*)

Institute for Materials Research, Tohoku University, Oarai, Ibaraki-ken 311-1313, Japan e-mail: konashi@imr. tohoku. ac. jp

T. Yokoyama

Toshiba Nuclear Engineering Services Corporation, 8 Shinsugita, Isogo-ku, Yokohama 235-8523,Japan

© The Author(s) 2015 169

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_16

FBRs provide high fast neutron flux, wherein the neutron reaction cross sections are small compared with those in the thermal energy region. Moderation of fast neutrons by hydride materials was considered to increase the transmutation rate [2—5]. In this chapter, enhancement of transmutation of MA by an MA-hydride target is studied. Target assemblies containing MA-hydrides are placed in the radial blanket region. Fast neutrons generated in the core region are moderated in the hydride target assembly and then produce high flux of thermal neutrons, which have large nuclear reaction cross sections to actinides. The MA-hydride target also has another advantage to load MA to limited space. The target of (MA, Zr)Hx increases mass of MA and hydrogen density in the blanket region compared with MA and ZrH16 loaded separately [4, 5].

Hydride fuels have been used in TRIGA reactors of General Atomics (GA) for many years [6]. On the other hand, hydride materials do not have much history of use in FBRs. Recently, a control rod of FBR with hafnium (Hf)-hydride has been studied [7]. In this chapter, the MA-hydride target pin was designed using exper­imental data of Hf-hydride.

The representative radioactive nuclides in this study (Table 20.5) include not only the important nuclides for various evaluations of radioactive wastes but also the nuclides whose concentrations have been measured in the past, which will be useful for the validation of numerical evaluations.

Target nuclides of sensitivity analyses were selected on the basis of two criteria. The first was that the concentrations of activation products be larger than or comparable to the concentration of fission products generated from impurity ura­nium in the materials. The contents of impurity uranium in Zircaloy-2 and SUS304 stainless steel were 0.00035 wt% and 0.0001 wt% [5], respectively. This value in INCONEL alloy is unknown. The second criterion was that the concentrations of activation products be comparatively large. In these analyses, activation products with concentrations more than 1 x 10~9 g/t were chosen.

The concentrations of activation products larger than 1 x 10~9 g/t and fission products generated from impurity uranium are shown in Table 20.6. The fission products were calculated under the condition that the initial composition contains only the uranium impurity. Table 20.6 also shows the selected target nuclides that satisfy the foregoing criterion: 17 nuclides in Zircaloy-2 and Zircaloy-4, 8 nuclides in SUS304 stainless steel, and 16 nuclides in INCONEL alloy were selected as the target nuclides of sensitivity analyses.

No. Nuclide Half-life No. Nuclide Half-life No. Nuclide Half-life 1 H-3 12.3 years 21 Se-79 377.0 E3 years 41 Cs-137 30.1 years 2 Be-10 1.5 E6 years 22 Rb-87 48.1 E9 years 42 Ba-133 10.5 years 3 C-14 5.7 E3 years 23 Sr-90 28.8 years 43 Sn-126 198.0 E3 years 4 Na-22 2.6 years 24 Nb-94 20.3 E3 years 44 Sb-124 60.2 days 5 Si-32 153.0 years 25 Nb-95 35.0 days 45 Sb-125 2.8 years 6 S-35 87.5 days 26 Mo-93 4.0 E3 years 46 Ce-141 32.5 days 7 Cl-36 301.0 E3 years 27 Mo-99 2.7 days 47 Ce-144 284.9 days 8 K-40 1.2 E9 years 28 Tc-99 211.1 E3 years 48 Eu-152 13.5 years 9 Sc-46 83.8 days 29 Tc-99m 6.0 hours 49 Eu-154 8.6 years 10 Ca-41 102.0 E3 years 30 Zr-93 1.5 E6 years 50 Eu-155 4.8 years 11 Cr-51 27.7 days 31 Zr-95 64.0 days 51 Gd-153 240.4 days 12 Mn-54 312.1 days 32 Ru-103 39.3 days 52 Ho-166m 1.2 E3 years 13 Fe-55 2.7 years 33 Ru-106 1.0 years 53 Hf-181 42.4 days 14 Fe-59 44.5 days 34 Ag-108m 438.0 years 54 Au-199 3.1 days 15 Co-57 271.7 days0 35 Ag-llOm 249.8 days 16 Co-58 70.9 days 36 1-129 15.7 E6 years 17 Co-60 5.3 years 37 Cd-113m 14.1 years 18 Ni-59 76.0 E3 years 38 1-131 8.0 days 19 Ni-63 100.1 years 39 Cs-134 2.1 years 20 Zn-65 244.1 days 40 Cs-135 2.3 E6 years aHalf-life is referred to ORLIBJ40 decay library bE3 years = 103 years, E6 years = 106 years °Half-life of Co-57 is not contained in ORLIBJ40, so it is referred to [4]

As clearly stated in the SCJ report, “management of the total amount” has two connotations: “setting an upper limit for the total amount” and “controlling increases of the total amount.” “Setting an upper limit for the total amount” corresponds to the withdrawal from nuclear power, and the level of upper limit depends on the tempo of that withdrawal. On the other hand, “controlling increases of the total amount” corresponds to keeping nuclear power in the future with strictly controlling increases of the total amount, and the amount of disposed waste per unit of generated power must be controlled to the smallest amount possible. There are many technical options to control the increase of the total amount of HLW, for example, increasing burn-up of fuels, transmutation of radioactive nuclides, and longer temporal storage of HLW, which secure time for radioactivity to decay.

However, in fact, many readers of the SCJ report mistakenly recognized that management of the total amount means setting an upper limit for the total amount, and thus believed that SCJ proposed withdrawal from nuclear power: this is a complete misunderstanding. At the background of the proposal of management of the total amount, there is recognition that we should respond to the concerns on the limitless increase of HLW.

To accurately calculate neutronics parameters, we have to use reliable calculation methods and nuclear data. For this purpose, we can use valuable measured data obtained from fast critical assemblies and fast reactors by applying the bias factor method [13] and the cross-section adjustment method [14]. In these two methods, it is necessary to consider that there are two kinds of errors, systematic and statistical errors, in measured and calculation errors. Here we propose a method to remove the systematic errors to improve prediction accuracy. Measured data Re have a system­atic error Reb and a statistical error Res and are expressed by

Re — Re0(1 + Reb + Res) (17.35)

where Re0 is the true value. Also, calculated neutronics parameters Rc are expressed

by

Rc — Rc0(1 + Rcb + Rcs + SAa) (17.36)

where Rc0 is the true value, Rcb is systematic error, Rcs is statistical error from calculation methods, and SAa is the error from cross-section error. To eliminate the systematic errors in measurements and calculations, we consider the ratio of measurement to calculation, called bias factors:

e _ 1 + Reb + Res

Rc 1 + Rcb ^ Rcs ^ SAa

Because the average of statistical errors becomes zero, the variance of f becomes

V (f) — V (Res) + V (Rcs) + SWST (17.38)

where W is the variance of nuclear data used. In deriving Eq. (17.38) it was assumed that all the systematic and statistical errors are smaller than unity and that there is no correlation between statistical errors of measurements and calculations. From Eq. (17.38) we can say that if there is no statistical error, the bias factorf is within the range of

1 — ca < f < 1 + ca a —л/W) (17.39)

with the confidence level of 65 %(c — 1), 95 % (c — 2), or 99 %(c — 3). Therefore, if f is outside the range, we can say in the foregoing confidence level, there is a systematic error of

Reb — Rcb — |1 — f I — ca (17.40)

For sodium void calculations, calculated values are the sum of positive nonleakage components and negative leakage components. The negative leakage components are difficult to estimate because the transport effect has to be consid­ered in calculating the neutron steaming. Therefore, there may be a nonnegligible systematic error in the leakage term RLb when the void pattern is leaky. By considering such a void pattern, we can discard the leakage term in systematic errors. Thus, we can determine the systematic errors. After the removal of the systematic errors, we can apply the cross-section adjustment method or the bias factor method to improve the calculation accuracy. In the cross-section adjustment method [13], the adjusted cross section is determined so as to minimize the functional J

J = (T — TojW-1 (T — Toy + R — Rc(T)][Ve + Vm]-1 [Re — Rc(T)}‘ (17.41)

where W is the cross-section covariance data, and Ve and Vm are the variance of measured data and calculation method, respectively. In Eq. (17.41), we replace Re and Rc(T) by

because the true values are unknown, they are approximated by Re and Rc. The adjusted cross section is given by

The covariance of the adjusted cross section is expressed by

W = W — WG[GWG’ + Ve + Vm]GW (17.44)

This expression is the same as in [13], but we have to use the adjusted cross section shown in Eq. (17.43).

Using the adjusted cross section, the MA transmutation rate can be estimated by

Rf) (t) = Rf) (T0) + G[6] (T — T0) (17.45)

where the superscript (2) indicates the MA transmutation rate of the target reactor. The variance, the uncertainty, of the MA transmutation rate is given by

17.5 Conclusion

To realize the harmonization of MA transmutation and sodium void reactivity, the MA transmutation fast reactor core concept, with an internal blanket between the MA-loaded core fuel region and the sodium plenum above the core fuel, was proposed. The feature of this core concept is that sodium void reactivity can be greatly reduced without spoiling core performance for normal operation.

To accurately evaluate neutronics parameters in a MA transmutation fast reactor, we improved the calculation methods for estimating MA transmutation rates and safety-related parameters such as sodium void reactivity. For the MA transmutation rate, we introduced a definition of MA transmutation for individual MA nuclides and a method for calculating the MA transmutation rates. To evaluate the prediction accuracy of neutronics parameters, we proposed a new method that can eliminate systematic errors of measurements and calculations, and introduced a method to reduce the prediction uncertainty based on the cross-section adjustment method or the bias factor method. Furthermore, we improved the sensitivity, which is neces­sary to evaluate the uncertainty, by considering the effect of self-shielding.

Acknowledgments A part of the present study is the result of “Study on minor actinide trans­mutation using Monju data” entrusted to University of Fukui by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).

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.

Hiroki Sono, Kotaro Tonoike, Kazuhiko Izawa, Takashi Kida, Fuyumi Kobayashi, Masato Sumiya, Hiroyuki Fukaya, Miki Umeda, Kazuhiko Ogawa, and Yoshinori Miyoshi

Abstract For the decommissioning of the Fukushima Daiichi Nuclear Power Stations, fuel debris involving molten structural materials should be retrieved from each reactor unit. The fuel debris, which is of uncertain chemical composition and physical state, needs to be treated with great care from the standpoint of criticality safety. For developing criticality control for the fuel debris, the Japan Atomic Energy Agency (JAEA) has been planning to modify the Static Experiment Critical Facility (STACY) and to pursue critical experiments on fuel debris. STACY, a facility using solution fuel, is to be converted into a thermal critical assembly using fuel rods and a light water moderator. A series of critical experiments will be conducted at the modified STACY using simulated fuel debris samples. The simulated fuel debris samples are to be manufactured by mixing uranium oxide and reactor structural materials with various chemical compositions. This report summarizes a facility development project for an experimental study on criticality control for fuel debris using the modified STACY and simulated fuel debris samples.

Keywords Critical facility • Criticality control • Criticality safety • Fuel debris • Fukushima Daiichi • Simulated fuel debris sample • STACY

22.1 Introduction

In the severe accident at the Fukushima Daiichi Nuclear Power Stations (NPS), most of the fuel loaded in the cores of Units 1, 2, and 3 was seriously damaged and melted, resulting in a considerable amount of fuel debris [1]. It is believed that some parts of the fuel debris involve molten structural materials such as zircaloy, stainless steel, and concrete. This fuel debris, which contains much burned fuel, still continues to emit radiation and heat. For decommissioning of the Fukushima Daiichi NPS, all the fuel debris should be retrieved from the pressure and contain­ment vessels of each reactor unit.

In preparation for retrieval of the fuel debris from the Fukushima Daiichi NPS, however, there remain the following serious problems: (1) leakage of cooling water from containment vessels, (2) inflow of groundwater into reactor buildings,

(3) maintenance of subcritical state of the fuel debris, and (4) shielding of radiation from the fuel debris [2]. The cooling water of the fuel debris concerns these four problems. In a similar accident that occurred at Three Mile Island NPS Unit 2 (TMI-2), where its pressure vessel was not seriously damaged, all these problems were settled or did not arise because the pressure vessel could be filled with cooling water containing highly concentrated boron as a neutron absorber and radiation shield [3]. In contrast, all four problems make it extremely difficult to retrieve the fuel debris from each reactor unit of the Fukushima Daiichi NPS.

The fuel debris, which has uncertain chemical composition and physical state, needs to be treated with great care from the aspect of criticality safety. In particular, large blocks of fuel debris can cause a change in physical state, such as size and water content, when they are broken into fragments to be retrieved in cooling water. Furthermore, a recent study on fuel debris resulting from the molten core-concrete interaction has revealed its potential for criticality [4]. There will probably be no risk of a criticality accident if it is possible to keep a high concentration of boron in the cooling water and take the criticality control measures that were used in the TMI-2 accident. However, these measures will be difficult unless both (1) the leakage of cooling water and (2) the inflow of groundwater are completely stopped. If not, retrieval of the fuel debris will require alternative approaches to criticality control in cooling water or dry retrieval with radiation shielding.

The authors focus on the former approach: new criticality control measures for fuel debris, such as criticality safety standards and criticality monitoring method­ology [5]. This report summarizes a facility development project for an experimen­tal study on criticality control for fuel debris.

Nobuyoshi Ishii, Shinichi Ogiyama, Shinji Sakurai, Keiko Tagami, and Shigeo Uchida

Abstract It has been recognized that carbon-14 (14C) is one of the dominant radionuclides affecting dose from transuranic (TRU) wastes. This radionuclide has a decay half-life of 5,730 years, and 14C organic materials have very low sorption properties to clay and rock in the environment, which raises some concerns about the releases of 14C to the biosphere from radioactive waste repositories. For the safety assessment of TRU waste disposal, we studied the behavior of 14C in rice paddy field soils. We also determined key parameters such as soil-soil solution distribution coefficients (Kds) and soil-to-rice plant transfer factors (TFs) of 14C in the field soils. The TFs were obtained in laboratory and field experiments. In our laboratory experiments, we used [1,2-14C] sodium acetate as a source of 14C because it has been suggested that low molecular weight organic-14C compounds are released from metallic TRU wastes. The results showed that 14C-bearing sodium acetate in irrigated paddy soils was rapidly decomposed by indigenous bacteria. Although some of the 14C was assimilated into the bacterial cells, most of the 14C was released into the air as gaseous compounds. The main chemical species of 14C gases was 14CO2, and a part of the released 14CO2 gas was used by rice plants during photosynthesis. Only a negligible amount of 14C was absorbed through the roots. Therefore, the contamination of rice plants is mainly caused by gasification of 14C, and microorganisms are responsible for driving this process. The activity of microorganisms is a key issue in the behavior of 14C in paddy fields.

Keywords Bacteria • Behavior • Degradation • Radiocarbon • Rice paddy fields • Safety assessment • TRU wastes

26.1 Introduction

Transuranic (TRU) wastes contain a variety of radionuclides, for example, Np, Pu, and long-lived radionuclides such as 14C and 129I. In Japan these wastes are categorized into four groups in accordance with their physical properties and the

N. Ishii (*) • S. Ogiyama • S. Sakurai • K. Tagami • S. Uchida

Research Center for Radiation Protection, National Institute of Radiological Sciences,

4-9-1 Anagawa, Inage-ku, Chiba 263-8555, Japan

e-mail: nobu@nirs. 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_26

concentration of radioactive materials. Group two waste includes hull and end piece wastes with relatively high amounts of 14C, and leaching of low molecular weight 14C organic materials from simulated hull wastes has been reported [1]. The 14C organic materials have very few sorption properties to clay and rock, and 14C has a relatively long half-life of 5,730 years. These properties raise concerns about releases of 14C to the biosphere from radioactive waste repositories.

Rice is a major agricultural crop throughout Asia, and thus human exposure to 14C through rice intake must be considered. To reduce the risk of the internal radiation dose from 14C, it is important to clarify the behavior of 14C in rice paddy fields. In this study, we determined transfer pathways of 14C through the rice paddy fields to rice grains. Environmental parameters such as soil-soil solution distribution coefficients (Kds) and soil-to-rice plant transfer factors (TFs) of 14C were also determined, because these parameters are often used in transfer models to predict the behavior of radionuclides in the environment. From a series of our experimental results, we describe the behavior of 14C in rice paddy field soils and the importance of microbial activity.