Toshikazu Takeda, Koji Fujimura, and Ryota Yamada

Abstract To effectively transmute minor actinides (MAs), which have long-lived radioactivity and high decay heat, fast reactors are very promising because many minor actinides can be loaded and transmutation rates are high compared to light water reactors. With the increase of loaded minor actinides, the neutron spectrum becomes hard and core safety parameters will deteriorate. Especially, the sodium void reactivity increases with MA addition to cores. To overcome the difficulty, we propose MA transmutation fast reactors using core concepts with a sodium plenum and internal blanket region in reactor cores. Therefore, cores become complex, and calculation accuracy becomes poor. To accurately evaluate the neutronic properties such as MA transmutation rate and sodium void reactivity, we improved calculation methods. In this chapter we show new methods for calculating MA transmutation rates for each MA nuclide, for calculating the uncertainty of MA transmutation using sensitivities. A new sensitivity is derived that is defined as a relative change of core parameters relative to infinite-dilution cross sections, not effective cross sections. To eliminate bias factors in estimating core parameter uncertainties, a new method is proposed. This method is used to reduce the calculation uncertainty through the use of adjusted cross sections.

Keywords Calculation methods • Fast reactors • Minor actinide • Sensitivity • Sodium void reactivity • Transmutation

17.1 Introduction

The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been acknowledged worldwide. However, acceptance of large-scale contributions would depend on satisfaction of key drivers to enhance sustainability in terms of economics, safety,

T. Takeda (*) • R. Yamada

Research Institute of Nuclear Engineering, University of Fukui, Fukui, Japan e-mail: t_takeda@u-fukui. ac. jp

K. Fujimura

Hitachi Works, Hitachi-GE Nuclear Energy, Ltd., Ibaraki, 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_17

adequacy of natural resources, waste reduction, nonproliferation, and public accep­tance. Fast reactors with fuel recycle enhance the sustainability indices signifi­cantly, leading to the focus on sodium-cooled fast reactors (SFR) in the Generation IV International Forum (GIF) and the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative of the International Atomic Energy Agency (IAEA).

The necessary condition for successful fast reactor deployment is the under­standing and assessment of innovative technological and design options, based on both past knowledge and experience, as well as on ongoing research and technology development efforts. The severe accident at Tokyo Electric Power Company’s Fukushima Dai-Ichi Nuclear Power Station caused by the Great East Japan Earth­quake and tsunami on March 11, 2011 prompted all countries to redefine their fast reactor programs. To achieve the successful deployment of fast reactors, drastic safety enhancement is the most important issue to be established, especially in Japan, where the restart of nuclear power plants once these have been stopped is a serious matter of argument.

The safety aspects of fast reactors (FRs) have been reviewed [1—4] in represen­tative countries that have developed or have a plan to develop fast reactors in the near future, especially after the Fukushima accident. These countries are improving the safety of SFRs by considering the DiD (defense in depth). The designs of SFRs should have tolerance to DBA (design basis accidents) and BDBA (beyond design basis accident) caused by internal and external events. The inherent safety and passive safety should be effectively utilized for reactor shutdown and reactor cooling. For the case of severe accidents, it is indispensable first to shut down the reactors. Furthermore, decay heat removal is also indispensable even in the case of SBO (station black out). For SFRs, natural circulation can be expected in the sodium heat transport systems and the decay heat can be removal to atmosphere by the air cooling system.

In Japan, the Ministry of Education, Culture, Sports, Science and Technology has launched a national project entitled “Technology development for the environ­mental burden reduction” in 2013. The present study is one of the studies adopted as the national project. The objective of the study is the efficient and safe transmuta­tion and volume reduction of MAs with long-lived radioactivity and high decay heat contained in HLW in sodium-cooled fast reactors. We are aiming to develop MA transmutation core concepts harmonizing MA transmutation performance with core safety. The core concept is shown in Chap. 2. Also, we are aiming to improve design accuracy related to MA transmutation performance. To validate and improve design accuracy of the high safety and high MA transmutation perfor­mance of SFR cores, we developed methods for calculating transmutation rates of individual MA nuclides and estimating the uncertainty of MA transmutation.

A new definition of transmutation rates of individual MA nuclides is derived in Chap. 3. Using the definition, one can understand the physical meanings of trans­mutation for individual MA nuclides. Sensitivities are required to estimate the uncertainty of MA transmutation rates from cross-section errors. In Chap. 4, sen­sitivity calculation methods are derived. First, the sensitivity calculation method relative to infinite-dilution cross sections is introduced. The MA transmutation rates are burn-up properties. Thus, the sensitivity calculation method for burn-up-depen­dent properties is derived. Finally, we investigate how many energy groups are required in sensitivity calculations. Calculated MA transmutation rates have large uncertainties resulting from the large uncertainties in MA cross sections. To reduce these uncertainties in MA transmutation rates, we introduce a new method to reduce prediction uncertainties of MA transmutation rates in Chap. 5. In this method, we eliminate bias factors included in experiments and calculations by using ratios of the calculation to the experiment of core performance parameters. After removing the bias factors, the cross section is adjusted using measured data. The conclusions are shown in Chap. 6.

Fuel assemblies with the design called “BWR STEP 3” had been loaded in the reactors. Each new fuel assembly contains six kinds of uranium dioxide (UO2) fuel (Fig. 21.1, Table 21.1). The most popular initial 235U/U enrichment in the fuels is

4.4 wt%, whose inventory per assembly is 76.8 kgU. The fuel of 9.6 kgU per assembly has the highest initial enrichment of 4.9 wt%. The initial uranium inventory in total is 170.9 kgU per assembly, including fuels of other enrichments and of the UO2-gadolinium oxide (Gd2O3) composite [4].

The Unit 1 reactor in 1FNPS had 400 assemblies, which consisted of six batches of burn-up. Each of the Unit 2 and 3 reactors had 548 assemblies of five batches. Among these assemblies, 64 in the Unit 1 reactor, and 116 in the Unit 2 reactor, had a low burn-up of only 3-5 GWD/t (Table 21.2). Other assemblies of the same number are older but still have a burn-up as low as 15-16 GWD/t. The oldest assemblies have a burn-up of about 40 GWD/t [5].

Assembly Pitch 15.24cm

U02 23SU 4.9wt% @ U02 235U 3.4wt%

2 U02 23SU 4.4wt% U02 235U 2.1 wt%

© U02 235U 3.9wt% (G) U02-Gd203 3.4wt%

w w 2 2 3 Gd203 5.0wt%

Fig. 21.1 Benchmark model of the BWR STEP3 fuel assembly

The condition of the fuel debris has not yet been identified in any reactor except estimations by severe accident analysis codes. Study of the TMI-2 fuel debris [6], however, suggests that various kinds of fuel debris may also be produced in the 1FNPS reactors, such as hard and loose debris. Especially, loose debris may show a wide variety of composition including structural materials such as Zircaloy and

Table 21.1 Initial uranium inventory in a boiling water reactor (BWR) STEP 3 fuel assembly

steel. Boron originating from the control rods cannot be expected necessarily to coexist with the fuel debris. It is also possible that the fuel debris in CVs has been generated through the molten core-concrete interaction (MCCI). It must be con­sidered that the fuel debris is not uniform and will be found at various locations.

The fuel debris is being cooled with nonborated water although it is highly preferable to add neutron poison and to maintain enough concentration in the water to secure the subcritical condition such as was performed after the TMI-2 accident. Boration is not realistic at present because of the coolant water leakage from CVs and underground water inflow to the coolant water circulation. Boron will be injected only in the event of re-criticality [7].

The founding of the All Japan Educational Debate Association, of which the author is a committee member, was the catalyst for a national debating contest, which was started in the Tokai, or central region, of Japan. This contest, having been sponsored for some years by Chubu Electric Power Company, debates energy-related issues. High school students have faced each other over topics such as “Japan should abandon nuclear power: for or against?”; “Television broadcasting time in Japan should be limited to save energy: for or against?” Although the topic of energy, as well as many other policy issues, needs to be thought about by the next generation, it is just such issues that the younger generation does not appear eager to tackle face on. This is where debate, with its game-like, competitive element can serve an important role. In the context of debate, young people have been shown to engage seriously with such issues.

With this in mind, the author organized a debate for her class of university students. They debated the following motion, suggested originally by Chiba University’s Assistant Professor, Daisuke Fujikawa: Japan should scrap the plan to store high-level radioactive waste underground: Do you agree or disagree? (1) This paper describes the procedure of the debate in the classroom, assesses its effectiveness, and discusses certain problems that emerged from this activity.

Kenji Nishihara, Kazufumi Tsujimoto, and Hiroyuki Oigawa

Abstract With consideration of the phase-out option from nuclear power (NP) utilization in Japan, an accelerator-driven system (ADS) for Pu transmutation has been designed and scenario analysis performed. The ADS is designed based on the existing ADS design for MA transmutation, and the six-batch ADS was selected as a reference design for scenario analysis. In the scenario analysis, the once — through scenario of light water reactor (LWR) spent fuel is referred to as a conventional scenario with a LWR-MOX utilization scenario. As the transmutation scenario, three cases of transmuters that are only-FR, only-ADS, and both-FR +ADS are analyzed. The numbers of necessary transmuters are obtained as 15 to 32 units, and the necessary period for transmutation as 180-240 years. The benefit on repository by reduction of Pu and MA is reduction of repository area by a factor of five and of decay time of toxicity by one order of magnitude. The FR+ADS scenario would be a modest solution, although the ADS scenario is preferable if rapid transmutation is required.

Keywords ADS • Phase-out scenario • Scenario study • Transmutation

19.1 Introduction

After the Fukushima-Daiichi accident, Japan started a discussion of nuclear power (NP) utilization including a “phase-out” option in addition to the usual scenario of utilizing plutonium by deploying fast breeder reactors (FBRs). In the phase-out option, construction of new plants is limited and dependency on NP will be gradually reduced. One of the reasons supporting the phase-out scenario is an ambiguous prospect of conducting underground disposal of radioactive wastes. Increase of wastes can be limited or even stopped in the phase-out scenario, but spent fuels (SFs) containing plutonium (Pu) and minor actinides (MAs) will remain as a legacy of NP. “Direct disposal” to the underground of SFs confined in canisters is considered as a strong option to treat this legacy, but Pu and MAs that exist in the

K. Nishihara (*) • K. Tsujimoto • H. Oigawa

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

Ibaraki 319-1195, Japan

e-mail: nishihara. kenji@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_19

underground can be utilized for nuclear weapons and can cause public dose in the very far future over several tens of thousands of years. Instead of direct disposal, transmutation of Pu and MAs (TRU, trans-uranic) has been studied in many countries for the purpose of eliminating them from the waste.

Transmutation can be performed by a “transmuter” that is dedicated for trans­mutation with the lesser role of electricity generation. It contains a fast reactor (FR) and an accelerator-driven system (ADS), which are fast neutron systems with metal coolant. FRs have been mainly developed as breeder reactors but they act as a burner reactor in the phase-out scenario. The burner reactor has no blanket region for breeding, larger Pu content, and shorter operation-cycle length [1]. The ADS has been designed as an MA transmuter with a smaller amount of Pu but changes to a Pu transmuter in this scenario.

In the present study, an ADS for Pu transmutation (Pu-ADS) is designed by neutronics calculation based on the ADS for MA transmutation (MA-ADS). In the original design for MA transmutation, drop of criticality during depletion is very small, and a long operation cycle is achieved because MAs behave as fertile material. In the Pu-ADS, criticality decreases much more rapidly and design modification is necessary.

After the design of the Pu-ADS, a scenario study is performed by a nuclear material balance (NMB) code that was developed by the authors. The following items are revealed by the study: accumulation of TRU in the LWR SFs, necessary number of transmuters, reduction of TRU, reduction of repository footprint, and radiotoxicity by transmuters.

In Sect. 19.2, calculation methods for neutronics design and scenario code are introduced. Section 19.3 provides the neutronics design and resulting ADS. Section 19.4 discusses assumptions and results of the scenario study. The results are concluded in Sect. 19.5.

For the decommissioning of the Fukushima Daiichi Nuclear Power Station Units 1, 2, and 3, research and development activities have been pursued to retrieve fuel debris from the pressure and containment vessels of each reactor unit. In prepara­tion for the retrieval, however, there remain serious problems concerning the cooling water of fuel debris from the aspect of criticality safety.

To study the new criticality control measures for the fuel debris, the Japan Atomic Energy Agency has carried forward a project to modify the Static Exper­iment Critical Facility (STACY) and to pursue critical experiments regarding the 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 in the modified STACY using simulated fuel debris samples. These samples are to be manufactured by mixing UO2 and reactor structural materials with various chemical compositions.

The license application for the STACY modification has been under safety review. 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 new criticality control measures applicable to the Fukushima Daiichi Nuclear Power Stations.

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.

Asako Shimada, Mayumi Ozawa, Yutaka Kameo, Takuyo Yasumatsu, Koji Nebashi, Takuya Niiyama, Shuhei Seki, Masatoshi Kajio, and Kuniaki Takahashi

Abstract The separation conditions for iodine species were investigated to analyze 129I in contaminated water and tree samples generated from the Fukushima Daiichi Nuclear Power Station (FDNPS). Inorganic iodine species in the samples from FDNPS were thought to be iodide (I_) and iodate (IO3~); therefore, the behaviors of these species during separation using a solid-phase extraction sorbent, Anion-SR, for water samples and combustion for tree sample were studied. When the amount of I was 1 pg and used within a few hours, I_ was extracted with the Anion-SR in 3 M NaOH and diluted HCl (pH 2) solutions, whereas IO3~ was only slightly extracted in these solutions. In contrast, 15 ng I_ with a larger amount of IO3~ (1 pg I) in the diluted HCl (pH 2) and allowed to stand for 1 day was only slightly recovered. It is possibly that I_ was changed to another species in a day in this condition. Iodate was successfully reduced to I_ with NaHSO3 in the diluted HCl solution and extracted with the Anion-SR. Consequently, the solution condition to analyze both I_ and IO3~ using Anion-SR was observed to be the diluted HCl at pH 2 with a reductant. For the tree samples, a combustion method was applied and

A. Shimada (*)

Nuclear Cycle Backend Directorate, Japan Atomic Energy Agency, 2-4 Tokai-mura, Naka-gun, Ibaraki, Japan

Fukushima Project Team, Japan Atomic Energy Agency,

2-4 Tokai-mura, Naka-gun, Ibaraki, Japan e-mail: shimada. asako@jaea. go. jp

M. Ozawa • Y. Kameo • T. Niiyama • S. Seki • M. Kajio Fukushima Project Team, Japan Atomic Energy Agency,

2-4 Tokai-mura, Naka-gun, Ibaraki, Japan

T. Yasumatsu • K. Nebashi

Tokyo Power Technology Ltd., 5-5-13 Toyosu, Eto-ku, Tokyo, Japan K. Takahashi

Nuclear Cycle Backend Directorate, Japan Atomic Energy Agency, 2-4 Tokai-mura, Naka-gun, Ibaraki, 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_27

the rate of temperature increase was optimized to avoid anomalous combustion. Greater than 90 % recovery was obtained for both I_ and IO3~, and the chemical species in the trap solutions was observed to contain I_.

Keywords 129I • Combustion • Fukushima Daiichi Nuclear Power Station • Iodine species • Isotopic exchange • Solid-phase extraction

27.1 Introduction

Because of the accident, a large amount of radioactive waste was generated at the Fukushima Daiichi Nuclear Power Plants (FDNPP). To establish the waste man­agement strategy, the radioactivity inventory has to be evaluated. Iodine-129 is one of the important nuclides of which the radioactivity has to be evaluated. Although I_ is considered a major species of 129I generated in the reactor, IO3~ and I2 are possibly generated, depending on the reactor conditions [1]. Furthermore, because seawater was introduced to the reactors for cooling down in the early phase of the accident and seawater contains the natural iodine species, 127IO3~, an isotope exchange reaction between IO3 and I may have occurred. Therefore,

analytical conditions to determine total I content, in this case IO3~ and I_, in water and tree samples were investigated in the current work.

Presently, contaminated water is accumulating in the basement of the reactor and turbine buildings at FDNPP. The accumulated water-processing equipment was installed to decontaminate and to desalinate. Consequently, secondary waste such as spent zeolite and sludge is generated. To evaluate the radioactivity inventory of the waste indirectly, water samples were collected from the inflow and outflow of the apparatus [2]. The contaminated water contains high levels of radioactivity of 137Cs, 90Sr, and other radionuclides. To limit radiation exposure of the analyst, rapid chemical separation of iodine species from these radionuclides is required. Chemical separation studies using the solid-phase extraction sorbent Anion-SR have been reported to rapidly separate I_ from major fission products such as Cs and Sr in contaminated water samples [3]. However, Anion-SR essentially extracts only I_ and not IO3_. Therefore, reduction of IO3~ to I_ is required to analyze total I. In this study, NaHSO3 was used as the reductant and the solution conditions were studied to reduce IO3~ to I_.

Because of the hydrogen explosion of FDNPP, trees on the site were contaminated by the radionuclides. Many of the trees were cut down to provide space to install tanks storing the contaminated water. Consequently, approximately 40,000 m3 of trees were stored in the site as radioactive waste [4]. A combustion method was used to analyze 129I in cement, ash, and soil samples [5,6]. To apply a combustion method for the tree samples, there were some subjects: evaporation and deposition of the organic materials and anomalous combustion. Therefore, decomposition of organic material to CO2 and H2O using oxidant was examined. In addition, the rate of temperature increase was controlled to avoid anomalous combustion. Furthermore, the influence of the chemical species, IO3~ or I_, on recovery was studied.

This chapter describes parametric analysis methodology and analysis results for Doppler feedback enhancement and burn-up reactivity swing reduction.

15.3.1 Parametric Analysis Methodology

A hypothetical 300 MWe fast reactor core was used for the parametric survey to enhance Doppler feedback and burn-up reactivity swing. Table 15.1 and Fig. 15.2 show the assumed core conditions and RZ geometry for parametric survey, respec­tively. The calculation methods were as follows. Core burn-up characteristics were analyzed with the burnup calculation code STANBRE [13]. Reactivity coefficients were analyzed using the diffusion calculation code DIF3D [14]. The effective cross sections used in these calculations were obtained by the cell calculation code SLAROM-UF [15], based upon 70 group cross sections from JENDL-4.0 [16] with a self-shielding factor table as a function of background cross section. This method for the production of the effective cross sections is considered to be adequate to take into account the influence of each diluting material upon the self-shielding effect of heavy isotopes for the parametric study. Concerning mate­rial compositions, a homogeneous model of fuel, diluent, and spectrum moderator was used.

To begin with, in the survey to improve Doppler feedback, 21 elements to enhance resonance absorption were evaluated as a diluent material for the TRU alloy: Cr, Mn, Fe, Ni, Nb, Mo, Tc, Ru, Rh, Pd, Nd, Sm, Gd, Tb, Dy, Er, Tm, Ta, W, Os, and Au. Moreover, the effect by neutron moderators such as BeO, 7Li2O, 11B4C (100 % enrichment of 11B was assumed), and ZrH2 were investigated to clarify the impact against Doppler feedback by neutron spectrum softening. To compare the Doppler effect enhancement of various diluent materials and neutron spectrum moderators in a simple manner, each material was hypothetically added to TRU-10wt%Zr alloy. The amount of each material added was adjusted case by case to maintain 1.0 of k-effective at the end of cycle.

Next, in the evaluation to decrease the burn-up swing, the effects of the measures taken to increase the fissile amount at the beginning of the cycle were studied. The effects on burn-up reactivity swing were evaluated by reducing the core height, installing B4C shield at core peripheral, and increasing the number of refueling batches, which all lead to increase of the fissile amount at the beginning of the cycle.

Last, reflecting the results obtained by the parameter surveys, an optimal uranium-free TRU metallic fuel core was specified, and its feasibility in light of Doppler feedback and burn-up swing was evaluated by core performance analysis.

In the FR+ADS scenario, MA content in FR is limited below 5 % with respect to design limit and the remaining MA is transmuted in the ADS. In the first generation of transmutation from 2050 to 2110, six FRs and three ADSs are deployed, then three FRs and two ADSs in the second generation, and two FRs and one ADS in the third generation are built (Fig. 19.9). In the fourth generation, only ADS is utilized as to reduce TRU rapidly. The total amount of Pu and MA is reduced to 20 and 10 t, respectively, excepting MA in vitrified waste.

Japan’s research and development program for HLW disposal started in 1976 (Fig. 24.1). The first progress report was released in 1992 by PNC (Power Reactor and Nuclear Fuel Development Corporation). PNC was reorganized as JNC (Japan Nuclear Fuel Cycle Development Institute) in 1998, then merged with JAERI (Japan Atomic Energy Research Institute) to be JAEA (Japan Atomic Energy Agency) in 2005).

In 1999, JNC released the second progress report, and more importantly, in 2000 the Specified Radioactive Waste Final Disposal Act (Final Disposal Act, hereinaf­ter) was legislated.

The process for the legislation of the Final Disposal Act is shown in Fig. 24.2. As shown here, the Special Panel on Disposal of High-Level Radioactive Waste formed under the Japan Atomic Energy Commission (AEC) played an important

NSC: “Requirements of Geological Environment

to Select PIAs of HLW Disposal” (Sep. 2002)

Fig. 24.1 Evolution of high-level radioactive waste (HLW) disposal in Japan (Modified from ANRE/METI and JAEA [1])

Fig. 24.2 Legislation of specific radioactive waste final disposal act (June 2000) (Private com­munication from NUMO on November 13, 2013)

role along with the second progress report of JNC to set the contents of the Final Disposal Act.

Under the act, geological disposal is chosen for HLW disposal, and NUMO (Nuclear Waste Management Organization of Japan) was established for implementing the final disposal of HLW.

Organizational structure and the roles of related organizations set by the Final Disposal Act are shown in Fig. 24.3. As shown here, METI (Ministry of Economy, Trade and Industry) decides a basic policy and supervises all related activities. Owners of nuclear power plants provide a waste fund, which is collected from the electricity tariff, and the fund management is done by RWMC (Radioactive Waste Management, Funding and Research Center), while implementation of HLW dis­posal including site selection is borne by NUMO.

According to the current final disposal plan (Fig. 24.1), site of the final HLW disposal is to be selected in the 2020s and the final disposal will start in the middle of 2030s.

The Final Disposal Act was amended in 2007 to include TRU (trans-uranium) waste as a second type of specified waste (first type is HLW canisters, vitrified waste) because TRU waste is also to be disposed by geological disposal technology.

Although open solicitation for volunteer municipalities was employed for site selection, there has been no case except for a failed attempt by Toyo Town in Kochi Prefecture in 2007. Taking into account the failed attempt, METI added another scheme by the government to invite municipalities. The difficult situation, however, has continued, and after the Fukushima accident, the difficulties are increasing greatly.

The estimated sustainable life period of the existing final disposal sites for munic­ipal solid wastes (MSW) in Japan was only 18 years as of the end of FY2008. Therefore, waste avoidance, waste volume reduction, and recycling of MSW have been a national policy. However, the Fukushima Daiichi Nuclear Power Plant (F1) accident has created an entirely new dimension in environmental pollution problems. Because waste incineration and water treatment are, by their nature, the processes that concentrate pollutants such as radioactive cesium (rad-Cs) in ashes and sludge, MSW containing high concentrations of rad-Cs are produced in some areas where high atmospheric deposition of rad-Cs occurred in the aftermath of the F1 accident. As a result, recycling of MSW as concrete material and compost has become difficult, and their reuse has been often prevented because of public opposition even when rad-Cs concentrations in the wastes are below the clearance level (100 Bq/kg). Most of the citizens in the affected area are in hard opposition to disposal of rad-Cs-containing wastes even if radioactivity of the wastes is below the governmental limit for their disposal in landfills with leachate collection systems (i. e., 8,000 Bq/kg of Cs-134 + Cs-137). As the result, treatment residues are now piling up in many treatment facilities in some area, which may eventually jeopar­dize the treatment itself and exert serious negative impacts to everyday life. For example, sewage facilities in Fukushima Prefecture stored 74,401 t of dewatered sludge, molten slug, and incinerator ashes as of May, 2014. Therefore, suitable technologies to reduce the volume of such wastes or to decontaminate rad-Cs at low cost are urgently required.

Private companies and agencies have been working on sludge volume reduction through drying combined with granule processing [1] with the purpose of alleviat­ing storage problems at treatment facilities. High-temperature combustion of sludge with an additive for controlling basicity of incineration material also proved effective in condensing rad-Cs in fly ash. The cost of this technique, however, was high and would be justified only when a very strong social need for sludge volume reduction exists [2]. Another tested technique in this regard is extraction of sewage by hot 0.1 M oxalic acid followed by recovery of the extracted rad-Cs by zeolite [3]. The cost of the oxalic acid method is considered acceptable for large — scale sewage treatment facilities, although waste volume reduction is dependent on the amount of zeolite necessary to remove Cs from the extract. The Cs distribution factor value (ml/g) reported for zeolite was a few thousand whereas the values for ferrocyanide (Fer) compounds determined by the in situ Fer coprecipitation method

were between 104 and106 [4]. Apparently, the use of the Fer coprecipitation technique for rad-Cs removal from waste extract is appropriate to maximize waste volume reduction.

On the other hand, there are concerns on the outcome of using Fer, especially regarding the radiological risk of generating concentrated waste regarding rad-Cs and the chemical hazard from Fer compounds.

The concentration of rad-Cs in insoluble Fer precipitate [QCs (Bq/kg)] generated by adding 0.1 mM potassium ferrocyanide (the concentration used in most of our experiments) to the waste extract can be estimated as follows:

r EM Qcs= 100 100 pVC0

Here r is percentage of rad-Cs removed from the extract of MSW by Fer technique, E is percentage of rad-Cs extracted from MSW with water or oxalic acid, M is weight (kg) of MSW extracted by V (l) of the solvent, p is weight of Fer precipitate formed per unit volume of the extract (kg/l), and C0 is rad-Cs concentration (Bq/kg) in original MSW. Assuming that r, E, M, V, and p are 95 %, 90 %, 1 kg, 2.5 l, and 35 x 10~6 kg/l, respectively (the values typically encountered in our on-site tests), QCs (Bq/kg) is 9,771 C0, implying that rad-Cs concentration in the Fer precipitate can be about four orders of magnitude higher than that in the original MSW. The designated wastes with rad-Cs concentration >100,000 Bq/kg are going to be sent to the interim storage facility in Fukushima Prefecture and the waste volume reduction is going to be carried out at the interim storage site before final disposal. The wastes with rad-Cs concentration lower than 100,000 Bq/kg are going to be disposed in a leachate-controlled landfill constructed by the national government or a conventional municipal landfill. The amount of designated wastes stored in 12 prefectures is 140,343 t as of December 31, 2013 [5], but most are less than

100.0 Bq/kg in rad-Cs concentration. The amount of designated waste exceeding

100.0 Bq/kg is predicted to be 9,0001 with rad-Cs concentration varying between

120.0 and 540,000 Bq/kg depending on the origin of the waste [6]. If the extraction of the waste followed by Fer coprecipitation was conducted for 9,000 t of designated waste >100,000 Bq/kg, and r, E, M, V, and p values were the same as discussed early in this paragraph, 790 kg of insoluble Fer waste with rad-Cs concentration 1.2 x 109-5.3 x 109 Bq/kg (total amounts of rad-Cs, 9.2 x 1011 to 4.2 x 1012 Bq) can be generated. By comparison, the content of rad-Cs in a piece of vitrified high-level radioactive waste (weight, 500 kg) can be as high as

4.8 x 1015 Bq [7], that is, three orders of magnitude higher than that from 9,000 t of highly contaminated designated waste. With appropriate instrumentation and management, it is possible to handle the rad-Cs concentrated waste resulting from the volume reduction of designated waste relatively safely.

The chemical risk of using Fer compounds to concentrate rad-Cs also requires attention. Although reagents such as oxalic acid that may be used for the extraction of rad-Cs are biodegradable and the degradation products are nontoxic, Fer com­pounds contain a cyano group within their structure, and are potentially more hazardous. Chemical toxicity of Fer compounds, especially that of ferric ferrocy — anide (Prussian blue, PB hereafter), in mammals has been studied extensively because PB is a decorporation drug to treat internal rad-Cs contamination for both humans and livestock animals [8]. Based on laboratory animal studies, human male volunteer studies, and the experience of actual administration of PB to people contaminated with 137Cs, it was concluded that PB is basically nontoxic. The history of the use of Na-Fer, Ca-Fer, and K-Fer as food additives also indicates that the toxicity of Fer compounds is low. More important is the risk pertinent to the long-term decomposition of Fer and possibility of free cyanide leaching from waste materials. For example, large amounts of Fer compounds have been generated in the coal and petroleum gas purifier used in gas production industries. The used purifier (containing ferric ferrocyanide) was often abandoned around coal pyrolysis plants, etc., and has caused the pollution of soil and groundwater. The problem is widespread: 1,310, 234, and 1,100 to 3,000 sites are known in Germany, Nether­lands, and the U. S., respectively [9]. In these sites, groundwater contained cyanide complexes such as Fer rather than free and more toxic CN~ or HCN, probably because Fer was decomposed rapidly only when exposed to daylight and the decomposition of Fer in the dark underground was very slow [10]. Laboratory experiments showed that Fer was eluted from soil at pH > 13 whereas ferricyanide (Fe(III)-CN complex) was easily eluted by freshwater [11]. If Fer is to be used to concentrate rad-Cs, the resulting cyanide complex-containing waste should be managed properly by avoiding exposure to daylight and alkaline reagent. It is also possible to decompose Fer thermally or chemically (e. g., United States Envi­ronmental Protection Agency [12]) before the final disposal, depending on the cost allowed for the treatment.

Although the use of Fer coprecipitation technique has to be evaluated from the environmental safety considerations, it is also necessary to know if the technique is applicable to the actual MSW treatment at all. MSW waste extracts contain high concentrations of multiple transition metals (Fe, Mn, Cu, Zn, and Ni), alkali metal ions (Na and K), NH4+, alkaline earth ions (Ca and Mg), and anions (F_, Cl~, SO42~, NO3~, and PO43~). Mixtures of various kinds of insoluble Fer-metal pre­cipitates can be formed in such solutions, and the substitution of alkali metals in the precipitate should also occur. Solubility of each Fer-metal compound as well as the reaction kinetics between Fer ion and each metal should influence the amount and chemical structure of Fer-metal precipitate thus formed. The types and concentra­tions of anions in the solution affect the efficiency of coagulation of colloidal Fer solid, and thus the solid-liquid separation. Therefore, the feasibility of the Fer coprecipitation technique has to be tested and validated before its application to the actual treatment.

The objective of this work is to identify the factors that are likely to govern Cs removal from MSW extracts by the Fer coprecipitation technique and to optimize coprecipitation conditions for Cs removal. As detailed information on chemical components in the extracts of rad-Cs contaminated wastes is hard to obtain, we first obtained and analyzed uncontaminated MSW extracts (i. e., do not contain rad-Cs from the F1 accident), applied Fer precipitation techniques to the uncontaminated MSW extracts, and then proceeded to rad-Cs-contaminated waste treatment.