MA transmutation core concepts are developed by considering the amount of MA loading and the safety-related core parameters. Increase of MA loading in the core of a SFR makes the amount of MA transmutation large, which may decrease long­term radiotoxicity and decay heat of MA. On the other hand, loading a large amount of MA into the core of a SFR increases the sodium void reactivity. Therefore, harmonization of MA transmutation and sodium void reactivity is a key issue in designing the core concepts. As an example, Fig. 17.1 shows the relationship of MA content and sodium void reactivity; when the MA content is about 10 %, the sodium void reactivity increases by about 1$.

A homogeneous MA-loaded core of 750 MWe was designed in the FaCT project [1, 2] (Fig. 17.2). The configuration of this core is a conventional homogeneous core and homogeneous MA loading into the core fuel increases sodium void reactivity. Therefore, MA content in the core fuel assembly is limited to less than about 5 wt%. On the other hand, the safety issue has become more and more important since the Fukushima Daiichi NPP accident. Further, low void reactivity SFR designing has been pursued in Russia and France [5]. In this study, the coexistence of enhanced MA transmutation and zero void reactivity, that is, the harmonization of MA transmutation and core safety, is set as an objective.

Hitachi proposed an axially heterogeneous core (AHC) concept with sodium plenum [6, 7]. It was clarified that an increase of flux level at the top of the core fuel caused by the presence of the internal blanket and decrease of the height of the inner core fuel greatly decreased sodium void reactivity. In the core concept, sodium void reactivity can be extremely reduced without disrupting core performance for normal operation. The difference in core configurations between the Hitachi AHC with sodium plenum proposed in FR ‘91 [6] and the ASTRID ACV [5], which has been recently studied in France, is that absorber material is loaded in the upper shield for the ASTRID ACV.

We are going to optimize the specifications of the core shown in Fig. 17.3 to realize the high MA transmutation and zero sodium void reactivity. Figure 17.4

Fig. 17.1

shows the axial distribution of coolant density and the density coefficient of that core. The sodium density coefficient (%Дк/кк’/Др) in the sodium plenum becomes positive. On the other hand, sodium density change (Дp) in the sodium plenum becomes negative because of the increase of coolant temperature for an accident such as ULOF (unprotected loss of flow accident). Therefore, net sodium void reactivity becomes negative.

The mechanism for reducing sodium void reactivity of the core is that the axial neutron leakage is largely enhanced with coolant voiding in the sodium plenum. It is known that the evaluated leakage component of sodium void reactivity with diffusion theory might be overestimated by about 50 %. Therefore, calculation

Fig. 17.3 Vertical view of axially heterogeneous core with sodium plenum

accuracy for sodium void reactivity of the core with a sodium plenum might be poor. Thus, we should consider change of the neutron spectrum in the heteroge­neous MA loaded core. Figure 17.5 shows the neutron spectra for MOX driver fuel without MA and 10 % MA-mixed fuels in transmutation target with and without Zr-Hx. The spectrum of 10 % MA-mixed fuel in transmutation target without Zr-Hx is slightly softer than that of the MOX driver fuel without MA, because the 10 % MA-mixed fuel in the transmutation target has no fissile plutonium but the MOX driver fuel includes 239Pu and 241Pu. The neutron spectrum of the moderator mixture target fuel is clearly softer than other fuels.

The criticality safety handbook shows the minimum critical masses of homoge­neous uranium-water mixtures, 36 and 53 kg, respectively for the 235U/U enrich­ments of 5 and 4 wt%. Mass control limits that can avoid criticality are also given for heterogeneous UO2-water composites, that is, 28 kg for the 5 wt% enrichment. Even for the 3 wt% enrichment, its mass limit is still 67 kg [8]. These numbers are small compared to the possible uranium inventory in each fuel assembly with low burn-up.

Fuel debris may exist as composites of UO2 and structural materials such as Zircaloy and steel in the pressure vessels (PVs). Zircaloy does not greatly affect the criticality characteristics of fuel debris because of its small neutron absorption cross
section, but the iron in steel may increase the critical mass of fuel debris because it has strong neutron absorption.

The MCCI product would be a composite of UO2 and concrete. The major content of concrete is silicon dioxide, which has also a small neutron absorption cross section and neutron moderation capability. The critical mass of the UO2- concrete composite has been evaluated as 400 kg for the fresh UO2 of 5 wt% 235U/U enrichment. For the fuel burned up to 12 GWD/t, the critical mass can be as small as 800 or 2,000 kg, depending on how the effect of fission products is considered. Only the water bonded in concrete is considered in the evaluation; therefore, the critical masses can be smaller when the MCCI product is submerged in the coolant water

[9] . The mass of 2,000 kg is equivalent to 12 fuel assemblies. It is also known that a certain cluster of 16 assemblies in the Unit 2 reactor has an average burn-up of about 14 GWD/t. Thus, this evaluation is not far from reality.

Before knowing the actual condition of fuel debris, it is possible to compute critical conditions. Such work has been already conducted for many years to produce a handbook or a database for criticality safety. It is easy to extend these standards to wider conditions such as UO2-steel composite or UO2-concrete composite. The computation will supply a new set of “criticality maps of fuel debris.” These maps will indicate (Fig. 21.2) subcritical and critical conditions, and supercritical conditions that would likely bring severe consequences. In Fig. 21.2, the horizontal line represents variation of composition, and the vertical line represents variation of geometry. Composition on the right has higher reactivity and smaller critical volume. On the left, the composition is certainly subcritical, which can be excluded from the criticality control.

The actual criticality situation will be assessed by placing onto the map the fuel debris condition revealed by observations or sample analyses. It is also necessary to study how the condition can move on this map from expected changes such as temperature drop in the fuel debris or geometry changes caused by retrieval work of fuel debris, etc.

In 2013, the author compared and examined debate-focused courses held in two universities located in Aichi, central Japan; details of the courses are as follows.

1. “Enshu 1” (Seminar 1), a first-semester course, was taught to third — and fourth — year students at Sugiyama Jogakuen University’s Department of Human Sciences.

2. “Introduction to debate,” a first-semester course, was taught to second — and third-year students at Aichi Shukutoku University’s Faculty of Global Culture and Communication.

3. “Introduction to debate,” this time a second-semester course, was taught to second — and third-year students at Aichi Shukutoku University’s Faculty of Global Culture and Communication.

By studying these classes, the author hoped, first, to elucidate how students understanding of the debate issue—the disposal of high-level radioactive waste— would be affected by the lessons; and, second, to assess the effectiveness of debating classes on issues related to natural science facing modern society. In addition, students completed a questionnaire, “Fundamental Literacy for Members of Society” (2). Through analysis of the results of this survey, the author sought to gain new insights into methodology to promote a deeper understanding among students of issues facing modern society.

The course as listed here had four main features. First, the theme of the debate was announced at the beginning of the course. Second, rather than having students choose the subject for debate, the topic was assigned to the students. The fact that the topic was a science-related one was the third feature of the course. Because the students were from a humanities/social sciences background, their basic knowledge of science was, on the whole, rather limited. Because there was some concern that students would not be able to cope with debate, efforts were made to deepen students’ understanding of the issues involved before the actual debating contest. For example, Hajimu Yamana, professor at the Kyoto University Research Reactor Institute (KURRI), and Tomohisa Kakefu of the Japan Science Foundation were invited as guest speakers, and students also visited the Mizunami Underground Research Laboratory and the visitor facilities at Hamaoka Nuclear Power Station.

The fourth feature of the course was, therefore, that students were not left to research the topic by themselves, but were supported by, for example, being given the opportunity to listen and talk to experts. In addition, there was an element of experiential learning incorporated into the course in the form of, for example, the visit to Hamaoka Nuclear Power Station just mentioned.

Several codes were combined for ADS design (Fig. 19.1) containing proton transport, neutron transport, cross-section preparation, and depletion. The PHITS code [2] was used for transportation of protons and neutrons above the energy boundary of 20 MeV. Transportation of neutrons slowing down less than 20 MeV is interrupted, and position, direction, and energy are stored in a cutoff file. This file is processed as to be readable by the PARTISN code [3], which is a neutron transport code with multi-group theory. A 73-group cross section is prepared by SLAROM [4] code with the JENDL4.0 [5] nuclear data library. A 1-group micro-cross section is calculated by multiplying the 73-group cross section to the 73-group flux from PARTISN. One-group micro-cross section and total flux is used in the ORIGEN2

Fig. 19.1 Calculation

codes

code [6] to obtain material change after depletion. The material composition from ORIGEN2 is processed by a fuel control program that simulates reprocessing and fuel fabrication with adjustment of MA content ratio.

Akito Arima

Abstract It is my great honor and pleasure to speak to you this morning on the occasion of the International Symposium on Nuclear Back-end Issues and the Role of Nuclear Transmutation Technology after the accident of TEPCO’s Fukushima Daiichi Nuclear Power Stations. I would like to thank the organizers, especially Professor Hirotake Moriyama and Professor Hajimu Yamana, for inviting me to this Symposium.

I believe that this Symposium is very important and well timed to solve urgent problems concerning nuclear back-end issues and to develop nuclear transmutation technology. I myself am a nuclear theoretical physicist and am ignorant of nuclear technology. However, I believe that nuclear energy is indispensable for the future of human beings and that nuclear engineering must be further developed.

My talk consists of the following four subjects:

1. Demand for primary energy and electricity is increasing year by year.

2. Global warming is becoming a more serious problem.

3. Development of renewable energy must be promoted. However, it will require sufficient resources of time and budget.

4. Human beings cannot avoid depending on nuclear energy as well as other energy resources that do not emit CO2.

5. Nuclear technology must be developed.

(a) The safety technology of nuclear energy has to be developed for the future.

(b) The technology for the back-end of the nuclear fuel cycle has to be enhanced. The site for final disposal of nuclear wastes has to be determined as soon as possible in Japan, which is a responsibility of the Central Government.

(c) The research and development of innovative technologies, such as accelerator-driven systems, must be promoted to encourage the progress of final disposal.

A. Arima (*)

Japan Radioisotope Association, 2-28-45, Honkomagome, Bunkyo-ku, Tokyo 113-8941, Japan e-mail: arima@musashi. 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_23

(d) Research and development of nuclear technologies for reactor decommissioning, safety technology, back-end, etc., must be promoted intensively through interna­tional cooperation.

Keywords Accelerator-driven system • Decommissioning • Final disposal • Nuclear back-end • Nuclear energy • Nuclear transmutation

27.2.1 Reagents

Ultrapure grade NaOH solution (3 M) was purchased from Kanto Kagaku. Ultrapure grade of tetramethyl ammonium hydroxide (TMAH) solution was pur­chased from Tama Chemicals. The other reagents were all analytical grade or higher and purchased from Wako Chemical. Empore solid extraction disks, Anion-SR, were purchased from 3 M. Standard solution of 129I (35.8 kBq/g 129I; chemical composition, 50 Fg/g NaI and 50 Fg/g Na2S2O3 in H2O) was purchased from AREVA and diluted with H2O before use. Potassium iodide (analytical grade,

99.9 %) and KIO3 (analytical standard material grade, 99.98 %) were purchased from Wako Pure Chemical Industries, and ‘I and ‘IO3 were prepared from these reagents, respectively, because the stable iodine isotope is only 127I.

The effects of measures taken to enhance Doppler feedback, that is, diluent and spectrum moderator, are evaluated in this section.

As shown in Fig. 15.3, 6 among 21 diluent materials are found to enhance Doppler feedback more than Zr, the typical metallic fuel alloy. Although Nb, Ni, W, Mo, Fe, and Cr have greater potential to enhance Doppler feedback than Zr, there are some deficiencies that cannot be ignored. First, the melting points of Pu-Ni alloy and Pu-Fe alloy are below 500 °C, which is too low for nuclear fuel

[17] . Second, the melting point of Pu-W alloy is too high to fabricate fuel by injection casting because the melting temperature of W itself is above 3,000 °C.

Third, the allowable contents of Mo and Nb in the metal fuel alloy are too small to enhance the Doppler coefficients, which are 5 wt% and 3 wt%, respectively, under the condition to maintain their melting temperatures below 1,200 °C to prevent Am vaporization during injection casting [18]. Finally, the number of past experiences with Cr, for example, irradiation testing of Pu-Cr alloy, is less than enough to employ it as a diluent material for uranium-free fuel. Consequently, Zr was chosen as the fuel diluent material.

Then, as shown in Fig. 15.4, the absolute value of the negative Doppler coeffi­cient remarkably increased by introducing a spectrum moderator such as BeO, 11B4C, or ZrH2. The adoption of ZrH2, however, may cause dissociation of hydro­gen upon accident. Besides, the usage of 11B4C is costly because almost 100 % enrichment of 11B is necessary to enhance Doppler feedback significantly. There­fore, BeO was selected as a moderator material for the uranium-free core.

One of the impacts on the repository by transmutation is reduction of potential radiotoxicity, which is defined as total ingestion dose of the waste. Because waste is isolated from the public in the underground in reality, such direct ingestion never occurs and it is considered to be hypothetical, but it can represent the potential danger of waste. This toxicity of waste can be compared to that of uranium ore consumed for electricity generation causing radioactive wastes. Figure 19.10 illus­trates those toxicities corresponding to whole operation of LWRs and transmuters. Consumed natural uranium is 370,000 t.

When wastes are generated, the toxicity becomes higher than corresponding uranium ore by three orders of magnitude. Fission products such as Sr and Cs are dominant in the early several hundreds of years, although actinides contribute to toxicity after that. Toxicity in the LWR-OT scenario decays to the level of uranium ore after 100,000 years. By reducing Pu in the LWR-PuT scenario, the decay time becomes shorter, to 70,000 years. In the transmutation scenarios, shortening of decay time depends on the remaining amount of TRU. The decay time is about 10,000 years in the ADS scenario in which the remaining TRU is approximately 30 t, including vitrified wastes. In comparison between the LWR-OT scenario and the ADS scenario, the amount of TRU is reduced by one order of magnitude, so toxicity is also reduced by same order. If MA in the vitrified wastes is retrievable, the amount of TRU will be reduced to around 10 t, which implies toxicity is reduced to 1/30 and the decay time is around 2,000 years. Thus, the impact on toxicity by transmutation is significantly affected by MA in the vitrified wastes. Early intro­duction of MA partitioning to the RRP and R&D for retrievability from the glass wastes is of importance in this aspect.

Another impact on the repository is reduction of repository size by partitioning and transmutation of heat-generating nuclides in the wastes. Repository size is represented by a repository footprint, which is defined as an area devoted for waste excluding aisles, ducts, utility area, surface facility, and other.

In the LWR-OT scenario, the footprint corresponding to 45,000 t spent fuel reaches almost 4 km2, which is double the typical repository design for the glass

Electoricity generation capacity (GWe)
MA inventory (t)
	1970 2000 2030 2060 2090 2120 > b 2150 2180 2210 2240 2270
O^roCO-fc-OlOT^ICDCOO
2300

waste corresponding to 40-year operation of the RRP, because the spent fuel assembly occupies more area and heat generation from the Pu in it also contributes.

In the LWR-PuT scenario, two kinds of waste form are produced: 37,000 glass waste forms containing FP and MA, and spent fuel assembly of MOX of 4,000 t. Each occupies 1.6 km2, and the total is 3.3 km2. Although an amount of MOX spent fuel is smaller than that of UO2 spent fuel in the LWR-OT scenario by a factor of 11, it contains more heat-generating actinides such as Am and Pu, and its footprint is significant.

In the early several hundreds of years, 90Sr and 137Cs, whose half-life is around 30 years, are dominant for the footprint. They are separated in the RRP after 2025 as well as MA in the transmutation scenarios. They are absorbed by adsorbents such as zeolite and calcined to the waste form. Because half-life is rather short and the repository footprint is almost proportional to heat generation, long-term storage of the calcined waste is effective [9]. After 300 years of storage, an accumulated layout for the TRU wastes that is low heat generating and with long-term radioac­tive wastes becomes available. The footprint of this layout is smaller by two orders of magnitude than a typical layout for the vitrified waste. After separating 90Sr and 137Cs, 241Am, whose half-life is 432.2 years, becomes dominant, but this nuclide is transmuted in the transmutation scenarios. Heat generation from other fission products that are vitrified quickly decays to the level of the TRU waste.

As result of the long-term storage and transmutation, the footprint becomes almost constant after 2025 (Fig. 19.11). The glass waste form that is produced before 2025 and contains MA occupies 0.5 km2. In the ADS scenario, partitioning and long-term storage of Sr and Cs in the wastes produced from reprocessing of ADS spent fuel is not assumed because the impact is small. As a result, the footprint gradually increases to 0.8 km2. Technologically, separation is possible in the reprocessing for ADS, and it will be applied if the increase becomes significant. Steps observed in 2230 and 2330 are caused by wastes of remaining TRU that will

Fig. 19.11 Repository footprint when wastes are produced be diluted to the glass waste, considering heat generation. The remaining TRU of the FR scenarios are more than that of the ADS scenario.

In the transmutation scenarios, the final footprint is around 0.8 km2, which is a fifth of the LWR-OT scenario. As is the case of radiotoxicity, the time of introduc­ing partitioning is significant because more than half of the repository is occupied by glass waste forms with MA.

Geological disposal is a globally common technology of HLW disposal for either vitrified HLW canisters or the spent nuclear fuel itself. Figure 24.4 shows the HLW disposal scheme employed in Japan, which incorporates the multi-barrier concept in the scheme. The first barrier is the vitrified HLW canister itself; the solubility of vitrified waste is very low and it is contained in a canister made of stainless steel. The second barrier is a thick package made of carbon steel, the third is a buffer made of bentonite and sand, and last, the multiply packaged waste is placed in stable host rock located deep underground.

Difficulty in securing the safety of HLW disposal comes from the requirement that risks associated with HLW disposal must be maintained below an acceptable level for a very long period, beyond 10,000 years. Whatever technical measures are taken, risks would remain. This is basically the same problem as the case of safety measures for severe accidents of nuclear power plants. The safety issue of HLW disposal, however, is more difficult because of the very long time period in which human intervention for maintaining safety cannot be expected.

The reaction of soluble Fer salts (K, Na, or H compounds) with metal (Fe, Cu, Zn, Ni, Cd, Mn, etc.) ions in solution produces insoluble metal-Fer complexes. Fer ion and metal can precipitate as such compounds as A2M3[Fe(CN)6]2‘ nH2O, A2MFe (CN)6‘ nH2O, M2Fe(CN)6‘ nH2O, or as mixtures of these, depending on concen­trations of alkali metal ions (designated as A+) and divalent transition metal ions (M2+) in the solution. The elemental composition and crystal structure of pre­cipitates also vary with the combination of soluble Fer salt (e. g., lithium Fer, sodium Fer, or potassium Fer) and transition metal salts (e. g., chloride, nitrate, or sulfate salts) used [13]. Some trivalent metals (e. g., Fe3+) also precipitate with Fer.

The insoluble Fer compounds preferentially incorporate Cs into their structure by multiple mechanisms such as ion exchange, isomorphic substitution, and adsorp­tion. Distribution of Cs to metal-Fer precipitates is known to vary depending on solution pH and chemical characteristics of Fer solids. The distribution coefficient values (ml/g) are in the range of 104 to 106 for K-Co-Fer, 105 to 106 for Na-Ni-Fer, 105 for Na-Cu-Fer, 104 for K-Cu-Co-Fer, and 103 for K-Zn-Fer and Zn-Fer [5]. The coefficient value for Cs with Zn-Fer is small compared to those with Fe-Fer, Cu-Fer, and Ni-Fer complexes, but the aforementioned distribution coefficient values could have been underestimated owing to insufficient removal of colloidal Fer solids from solution. Alkali metal (i. e., principal constituents such as Na and K in solution) substitution in metal-Fer complexes also results in changes in Cs distribution [14]. The pH ranges for Cs distribution to preformed Fer complexes of Ni(II), Zn(II), Cu(II), and Fe(III) are reported to be 0 to 10, 1 to 8, 0 to 8, and 0 to 6, respectively [14]. Variations in Cs distribution to solid Fer within these pH ranges are also reported [15]. It is known as well that Fer should not be used in highly caustic and acidic solutions, because it is chemically decomposed in these reaction conditions.

The size of Fer precipitates is important as it determines settling velocity, which is a crucial factor for separation of Cs-containing solids in solution. Iron-Fer pre­cipitates are often found as colloidal particles, and their separation by gravity settling method is difficult whereas Ni-Fer precipitate settles more easily. The physical properties of precipitates depend on preparation procedures. For example, in strongly oversaturated solutions, very fine crystalline particles with a disordered lattice and higher solubility are formed incipiently, whereas an inactive solid phase is formed in slightly oversaturated solutions [16].

Solubility of metal-Fer precipitate is also a governing factor for Cs removal. The reported solubility product values for pure Fe4[Fe(CN)6]3, Zn2[Fe(CN)6], Cu2[Fe (CN)6], and Ni2[Fe(CN)6] are 3.3 x 10~41,4 x 10~16,1.3 x 10~16, and 1.3 x 10~15, respectively [17]. In the case of Ni2[Fe(CN)6], Fer concentration in solution should be higher than 10~5 M for insoluble Fer precipitate to be formed. Because the solubility of fresh metal-Fer compounds can be higher than that of pure compounds, the minimum Fer concentration used in this study was 10~4 M.

There are three different ways to use Fer to remove Cs: (1) addition of soluble Fer salts and metal elements to waste solution (in situ formation of Fer solid), (2) addition of freshly prepared insoluble Fer-metal complex slurry to waste solution, and (3) use of Fer-metal adsorbents in solution. The distribution of Cs to insoluble Fer compounds is highest when the in situ Fer formation method is applied. If appropriately used, the in situ method is the best for both decontamina­tion and waste volume reduction.