Category Archives: Study on Neutron Spectrum of Pulsed Neutron Reactor

Transmutation Calculation

Transmutation performances of the hydride MA target and related core features in FBRs have been evaluated with the method shown in Table 16.2. A three­dimensional continuous energy Monte Carlo Code MVP [9] and MVP-BURN [10] are used as burn-up calculations for evaluating the transmutation of MAs. The cross-section library applied in the calculations is JENDL-4.0, which is processed to be adjusted to the MVP code. In the burn-up calculation, the prediction-correction method is employed to improve accuracy with millions of neutron histories for the criticality calculation, where the accuracy of Eigen value is about 0.04 %.

Table 16.2 Calculation method for MA transmutation






Three-dimensional continuation energy Monte Carlo analysis code; MVP (burn-up routine is MVP-BURN)

1,200,000 neutron histories with 120 batches. Initial 20 batches are run to establish the initial neutron source distribution

Nuclear data

JENDL-4.0 library



Pin heterogeneous model

Table 16.3 Comparison of reduction ratio of MAs


Loading mass (kg)

Reduction mass (kg/year)

Reduction ratio after 1 year

Effective half life (year)




91.1 (33.0)a






27.6 (9.4)a



Ratio: Case1/ Case2





aValues in parentheses are reduction masses by fissions

Calculations have been done for two kinds of transmutation target. In case 1, the transmutation target was the MA hydride of (MA01, Zr09)H16. Calculation with metal MA0.1Zr0.9 target without H was done in case 2. The results of calculations are summarized in Table 16.3, where effective half-life is defined as the time such that the residual amount of MA is decreased to half of the MA loaded during the burn-up. The effective half-life is calculated to be 2.19 years in case 1 and

8.7 years in case 2, mainly because of the softened spectrum effect induced by the MA-hydride. The transmutation rate of the MA-hydride target is about three times higher than that of the MA-metal target. Figure 16.4 shows the change of total MA and each element of MA in the MA-hydride target with increase of time. Major elements in MA, that is, Np and Am are decreased simultaneously during the burn — up. The contribution of long-lived Cm (245Cm and 246Cm) is much smaller than that of Np and Am. The change of total MA in the MA-metal target is also shown in Fig. 16.4 for comparison.

The major mode of the transmutation in the present method is not fission but neutron capture (see Table 16.3). As shown in Fig. 16.5, Am and Np are mainly transmuted to Pu because of neutron capture, beta decay, and alpha decay. Recycled Pu is used as a driver fuel in this reactor.


Fig. 16.4 Change of each element in MA assemblies


242Cm *


* 244Cm *


243Cm —n;—










242Am 0





— ГГ7







* 240Pu 0






















Fig. 16.5 Chain transmutations for actinide nuclides




This study shows the sensitivity analyses of initial compositions and cross sections for activation products of in-core structure materials. The results clarified the source elements and nuclear reactions dominating the generation pathways of the activa­tion products even for the nuclides generated through complicated pathways. The sensitivity coefficients of initial compositions are beneficial for the evaluation of the error propagated from the uncertainty of the initial composition of target materials. The sensitivity coefficients of cross sections are effective in selecting
the objectives of nuclear reactions for the improvement of nuclear data. These results will contribute to improvement of the accuracy of numerical evaluations for the concentration of activation products.

The methodology of sensitivity analyses stated in this study is efficient for acquiring information about important impurity elements and nuclear reactions to evaluate the activation product concentrations. This methodology can be applied to the activations of ex-core structure materials if the appropriate one-group cross sections are prepared with a corresponding neutron spectrum.

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.


25.1.1 The Situation Now

Although the problems surrounding nuclear power have certainly become a social issue, it cannot be said that the present discussion on this issue is always calm and rooted in science. Elsewhere in the world, including in the USA and various European countries, the importance of public debate has been emphasized and reports published on the efficacy of specific examples. In Japan, however, partic­ularly since the Fukushima Daiichi nuclear accident, sensational media reporting

A. Yoshida (*)

Sugiyama Jogakuen University, 3-2005 Takenoyama, Nissin-City, Aichi, Japan e-mail: ayoshida@sugiyama-u. ac. 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_25

has not served to encourage sound debate on the issue of nuclear power. Moreover, this is not limited to the issue of nuclear power, nor is it a phenomenon that dates from the Fukushima disaster: issues around food safety and gender equality have been similarly characterized by emotive media reporting. Debate around these issues has not been informed by current scientific knowledge; indeed, what has prevailed is argument based on emotion engaged in without even an understanding of the relevant laws.

Being a democratic society entails that the path society follows is set according to the wishes of its citizens. These citizens exchange their divergent views and do their utmost to reach a consensus. If an agreement cannot be reached, society acts in accordance with the opinion of the majority. Necessary for such a process is that people think about an issue and express their views. In modern society, however, because many issues are complicated and difficult for people to adequately under­stand, often such issues are left to the “experts.”

Although the citizens’ right of self-determination should go hand in hand with responsibility, entrusting decision making to the experts has resulted in responsi­bility for these decisions being thrust upon them. Entrusting all responsibility to the experts could be expected to expedite decision making, but this has not been the case; instead, the emotional response of the public has contributed further to the deferring of decision making.

Th-MOX Fuels Irradiated in LWR Conditions

Within the European Framework Programmes, the study of Th fuels behavior in LWRs was first aimed at comparing the behavior and the applicability of various matrices to be used for the transmutation of Pu and minor actinides (projects THORIUM CYCLE, LWR-DEPUTY, OMICO). Comparisons were made with standard fuels (UO2, MOX), and also with so-called inert matrices fuels (using, for example, Mo or MgO as matrix in CERMET and CERCER fuel types, respec­tively). As explained earlier, irradiation experiments were performed in three facilities, namely, the KWO PWR, HFR, and BR2 Material Test Reactors.

The THORIUM CYCLE project was a 4-year project with the following partic­ipants: the coordinator NRG (NL), BNFL (UK), CEA (F), FZK and KWO (D), and JRC-IE and JRC-ITU (EU). The goals of this project, which started on 1 October 2000, were to supply key data for application of the Th cycle in LWRs. In particular, it included the study of

• The behavior of Th-based fuel at extended burn-up through an irradiation experiment of four short fuel pins [UO2, (U, Pu)O2, ThO2, and (Th, Pu)O2] up to 55 GWd/tHM in HFR, and an irradiation experiment of one short fuel pin [(Th, Pu)O2] to 38 GWd/tHM in a PWR (KWO); it should be noted that a previous irradiation of (Th, Pu)O2 in Germany (Lingen) achieved a burn-up of 20 GWd/tHM [7];

• The core calculations for Th-based fuel, including code-to-code validation, sensitivity check for significant isotopes 232Th and 233U, and the calculation up to 80-100 GWd/tHM for Th-MOX fuel.

The irradiation test in KWO enabled the investigation of the operational safety of Th-MOX rod behavior under realistic pressurized water reactor (PWR) condi­tions. The short test rod was inserted in a MOX assembly to provide the most realistic boundary conditions possible. The foreseen MOX carrier assembly had already been irradiated for one cycle. The cladding appeared in good condition after irradiation, and its creep-down, measured at the reactor site during the shut-down periods, as well as its general behavior, were well within the bounds of experience for UO2 fuels. The fission gas (Xe and Kr) release was about 0.5 % [8], which is about half that for equivalent MOX fuels at the same burn-up, but the linear power was lower than in equivalent U-MOX studies. Taking into account experimental uncertainties, the fuel behavior seems to be at least as good as U-MOX.

The THORIUM CYCLE project was completed in 2006, but the postirradiation experiments were performed under a subsequent experiment called LWR-DEPUTY (coordinator, SCK. CEN). In this program, the main tests on Th-MOX consisted of additional fuels studies (microscopy, radial distributions of elements and isotopes) and radiochemical analyses. The objective of these analyses was to obtain a reliable experimental database for burn-up analysis and to evaluate changes in the heavy nuclide content:

• To optimize the dissolution and analysis strategies

• To establish the first dataset on heavy nuclide and fission product content in irradiated Th-MOX to assess the overall uncertainties

• To use this dataset in a benchmark analysis program

The OMICO Project [4] was conducted from 2001 to 2007. Its scope included the study and modeling of the influence of microstructure and matrix composition on Th-MOX fuel in-pile behavior in normal PWR conditions. The following tasks were undertaken:

• Fabrication of the Th-MOX fuels at the JRC-ITU

• Irradiation in the “CALLISTO” PWR loop in BR2, representing real PWR conditions; the burn-up achieved at the end of this project was about 13 GWd/ tHM

• Nondestructive examinations (gamma-spectrometry, visual examinations) and microstructure studies

It should be noted that the pins were instrumented for pressure and fuel temper­ature determination. The test matrix was such that the Th-MOX could be compared with U-MOX and UO2 fuels. Another test parameter consisted of the fabrication process (homogeneous versus heterogeneous powder mixtures). The results of the temperature/pressure readings were primarily used to benchmark computer code models for Th-MOX fuels behavior in the first stage of their life.

Besides the irradiation, fuel characterization was performed, including thermal diffusivity measurements, and the results were published [4, 9]. The results show a similar thermal conductivity for (nonirradiated) Th-MOX as compared to U-MOX.

In the LWR-DEPUTY [5] project, selected samples of the OMICO and THO­RIUM CYCLE programs were extensively studied to provide experimental datasets suitable for evaluating their in-pile performance. The experimental data were the basis of a benchmark exercise on the Th-MOX fuel pin irradiated at the NPP KWO to investigate the qualification of the numerical tools and software packages. A scoping study of the leaching behavior was also conducted. In addition to the experimental work, steady-state and transient analyses were performed for different PWR designs fueled completely or partially with Th-MOX fuel. An assessment of steady-state parameters (reactivity, shutdown margin, and reactivity feedback coef­ficients) has been performed in comparison with UO2. All feedback coefficients are favorable for a safe operation under steady-state conditions. A comparative analysis of control rod ejection scenarios has also been performed, and it was found that the maximum values obtained for fuel and clad temperature and maximum fuel enthalpy are in line with the acceptance criteria for the current generation PWRs.

After 10 years of research sponsored through the EURATOM programs, the following conclusions can be drawn regarding the behavior of Th-MOX fuel in LWR conditions:

• Th-MOX has great potential and its fabrication as an oxide fuel is feasible

• Even at a laboratory-scale production route, Th-MOX shows a good in-pile performance

• Know-how on Th-MOX has increased, but

• Fuel performance obviously needs to be further improved before code calcula­tions can predict specific Th-MOX behavior

As a general conclusion, the results of these experiments have shown that Th-MOX behaves in a comparable way (even better in some aspects) to MOX, and that licensing Th-MOX in a LWR should not be problematic, although more experimental data on fuels representative of the future commercial fuels would be needed. Experimental data also demonstrate that Th fuels will be more resistant to corrosion than U fuels in the case of spent fuel geological disposal.

Manufacturing and Analytical Equipment for Simulated Fuel Debris Samples [12]

The simulated fuel debris samples (sintered pellets) are to be manufactured by mixing UO2 and reactor structural materials (Zr, Fe, Si, Gd, B, etc.) with various chemical compositions. These debris materials will be mixed in the form of oxide

powders. The manufacturing equipment for the debris samples is composed of a ball mill, compacting machine, and sintering furnace. The debris samples will be analyzed destructively or nondestructively to determine nuclide composition, O/U ratio, density, and impurities. The manufacturing ability is to be 300 pellets a month. The analytical precision is still a matter under consideration. The manufacturing and analytical equipment are to be installed in glove boxes in the experimental building adjoining the modified STACY.

Transfer of 14C from Soil to Rice Plants

Soil-to-rice plant transfer factors (TFs) of 14C, which was defined as 14C concen­tration in rice grains (Bq/kg-dry) divided by that in soil (Bq/kg-dry), were deter­mined by laboratory and field experiments. In the laboratory experiment using a

Partitioning ratio (%)


Solid phase

Liquid phase

Gas phase









Glutaraldehyde exposure




Cycloheximide exposure




Table 26.1 The partitioning ratios of 14C into solid, liquid, and gas phases for each treatment.

Fig. 26.4 Colonies of bacteria (a) and their autoradiography image (b). Heterotrophic bacteria have the ability to uptake 14C from an agar medium

growth chamber, we grew rice plants with addition of [1,2-14C] sodium acetate. This 14C compound was supplied once to rice plants in the flooding water just before blooming, and TF of 6.8 ± 2.4 on average was obtained. In these tracer experiments, rice plants were also cultivated without [1,2-14C] sodium acetate as negative controls in the same growth chamber as the 14C-treated rice. Interestingly 14C was detected even from the rice grains of negative control samples. These results suggested that the 14C-bearing gas, which was released from bacterial cells in rice paddy soils, was fixed by the rice plants in the negative controls through photosynthesis.


We also examined the possibility of root uptake of 14C by stable isotope techniques under field conditions [4]. If plant carbon originates from the atmo­spheric CO2, the 513C values in crops can be calculated using the 513C value, —8 %o in air [5], and the 13C fractionation ratio in photosynthesis by rice plants of —18 to —20 %o [6, 7]. The calculated 513C values in our study ranged from —28 %o to —26 %o, and the results implied that no soil carbon contribution occurred for white rice; however, by setting some conditions, for example, 13C fractionation ratio of 19%, we obtained the average TF value of 0.11 ± 0.04 for white rice. To compare these TF values obtained in laboratory and field experiments, it is necessary to pay attention to the difference between [1,2-14C] sodium acetate and the actual organic compounds present in the natural soil.



Подпись: Gas14C02 gas

Подпись: Release

Подпись: 4C-NaOAc image415 Подпись: Liquid


Подпись: Uptake

image418 Подпись: - Подпись: Bacteria Подпись: Sedimentation


Fig. 26.5 A conceptual diagram for the behavior of 14C in rice paddy fields

Development of Uranium-Free TRU Metallic Fuel Fast Reactor Core

Kyoko Ishii, Mitsuaki Yamaoka, Yasuyuki Moriki, Takashi Oomori,

Yasushi Tsuboi, Kazuo Arie, and Masatoshi Kawashima

Abstract A TRU-burning fast reactor cycle associated with a uranium-free trans­uranium (TRU) metallic fuel core is one of the solutions for radioactive waste management issue. Use of TRU metallic fuel without uranium makes it possible to maximize the TRU transmutation rate in comparison with uranium and plutonium mixed-oxide fuel because it prevents the fuel itself from producing new plutonium and minor actinides, and furthermore because metallic fuel has much smaller capture-to-fission ratios of TRU than those of mixed-oxide fuel. Also, adoption of metallic fuel enables recycling system to be less challenging, even for uranium-free fuel, because a conventional scheme of fuel recycling by electrorefining and injection casting is applicable.

There are some issues, however, associated with a uranium-free TRU metallic fuel core: decrease in negative Doppler reactivity coefficient from the absence of uranium-238, which has the ability to absorb neutrons at elevated temperatures, increase in burn-up swing, because fissile decreases monotonically in uranium-free core, and so on. The purpose of this paper is to evaluate the feasibility of the uranium-free TRU metallic fuel core by investigating the effect of measures taken to enhance Doppler reactivity feedback and to reduce burn-up swing. The results show a TRU-burning fast reactor cycle using uranium-free TRU metallic fuel is viable from the aforementioned points of view because the introduction of diluent Zr alloy, spectrum moderator BeO, and lower core height enables Doppler reactiv­ity coefficient and burn-up reactivity swing of uranium-free TRU metallic fuel to be as practicable as those of conventional fuel containing uranium.

Keywords Burn-up swing • Doppler reactivity feedback • Fast reactor • Metallic fuel • Trans-uranium • Uranium-free

K. Ishii • M. Yamaoka • Y. Moriki • T. Oomori • Y. Tsuboi • K. Arie (*) Toshiba Corporation, 8 Shinsugita-Cho, Isogo-Ku, Yokohama 235-8523, Japan e-mail: kazuo. arie@toshiba. co. jp

M. Kawashima

Toshiba Nuclear Engineering Service Corporation, 8 Shinsugita-Cho, Isogo-Ku, Yokohama 235-8523, 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_15

15.1 Introduction

For sustainable nuclear power deployment, not only ensuring its enhanced safety but also reduction of the environmental burden associated with radioactive waste management is a challenging issue for the international community. History has shown that obtaining public support is difficult for waste management plans that involve mass disposal of radioactive waste with a half-life of tens of thousands of years. Therefore, as one of the solutions, Toshiba has been developing a system that takes into account that, for the time being, light water reactors (LWRs) have a leading role in commercial nuclear power plants, which enables toxicity and radioactivity of high-level waste to be reduced to those of natural uranium within a few hundred years. This system is mainly characterized by a fast reactor core that does not contain uranium in its fuel, that is, uranium-free TRU fuel. The use of uranium-free TRU fuel makes it possible to maximize the TRU transmutation rate in comparison with fuel containing uranium because it prevents the fuel itself from producing new plutonium and minor actinides.

Although there was much research focused on TRU transmutation with uranium — free fuels, each of these seems to have drawbacks from some aspect. First, for instance, candidates such as Tc-based and W-based oxide fuel, inert matrix fuel such as the rock-like oxide fuel containing mineral-like compounds, and MgO-based oxide fuel provide solutions against issues associated with uranium — free operation, that is, decrease in Doppler reactivity feedback and increase in sodium void reactivity [13], but such types of inert matrix fuel may require new technologies for reprocessing. Additionally, many processing phases necessary for fabrication are costly. Second, an accelerator-driven transmutation system coupled with a fast reactor using uranium-free metallic fuel is another candidate that also can relax the issue of the reduced Doppler effect owing to its subcritical system [4­7], but installation of the accelerator facility at a fast reactor site is less cost competitive, especially when the system is not only a TRU burner but also a commercial power plant. Thus, it is worthwhile to develop the TRU transmutation system with uranium-free TRU fuel from the aspect of technological maturity and simplicity, which results in lower cost. Subsequently, the concepts for the TRU burner system with uranium-free TRU are derived from this background: fewer R&D needs and a simple system.

First, by contrast with inert matrix fuels, metallic fuel can be fabricated by the well-known injection casting method [8]. Moreover, metallic fuel is compatible with pyro-process reprocessing that has been developed since the 1960s [9]. Appli­cation of an accelerator-driven system for transmutation needs further R&D than that of a fast reactor system. Thus, the metallic fuel fast reactor is preferred for the system.

Second, we aim to develop the TRU-burning system in commercial power reactors while avoiding cost impact. For this reason, a system that can employ the pyro-process for fuel reprocessing would be preferable because it does not need


Fig. 15.1 Configuration diagram of the system to reduce nuclear waste burden

complex processes. Therefore, we introduce a metal fuel alloy that can be simply fabricated by injection casting and reprocessed by pyro-processing.

Additionally, in terms of reduction of nuclear waste burden, a metallic fuel fast reactor cycle has the great potential to transmute long-lived fission products (LLFPs) because of its excellent neutron economy [10, 11]. Moreover, it has an advantage for long-term energy security because the basic technology of the metallic fuel fast reactor cycle is also applicable to the future sustainable nuclear energy supply system.

For these reasons, Toshiba is developing a system to reduce nuclear waste burden using a TRU burner as shown in Fig. 15.1. The system is characterized by a closed fuel cycle that encompasses the following main facilities: fuel manufacturing plant to fabricate uranium-free TRU metallic fuel and LLFPs target from TRU and LLFPs extracted from LWR spent fuel, a fast reactor to burn those fuels, and recycling facilities to reprocess and refabricate the spent fuel from the fast reactor by pyro — processing. Although substances remain after reprocessing that must finally be disposed outside the cycle, their toxicity and radioactivity are diminished to the same level as those of natural uranium by enhancing burning and processing rates and storing them for a few hundred years within the system. Among the aforemen­tioned facilities in the system, this study focuses on the TRU-burning fast reactor and investigates the practicability of the uranium-free TRU metallic fuel core.

Result of FR

In the FR scenario as well as other transmutation scenarios, Pu from the RRP is at first fabricated as LWR-MOX fuel and burned in LWR. Pu is co-extracted with same content of U in the current RRP, although MA is vitrified as waste. MA partitioning is assumed to be introduced in 2025 and stored until 2045. In 2045, before introduction of transmuters in 2050, reprocessing of LWR-MOX spent fuel will begin and provide Pu to the transmuters.

FRs are to be introduced in 2050 when 250 t plutonium and 100 t MA remains. MA of 20 t is vitrified by the RRP before 2025 and is not available for transmuta­tion. Available TRU is 330 t. The required TRU to introduce an FR is approxi­mately 25 t, if we assume 41 % of Pu content and 15 % of MA content and employ 45.1 t from Table 19.10. Theoretically, 14 (=350/25) FRs can be introduced in 2050, but only 8 can be deployed in practice because the plant life of an FR is assumed to be 60 years and sufficient TRU must be kept until 2110. Available TRU gradually decreases to 2001 in 2110 by transmutation. After 2110, FRs are replaced and reduced to 3 units corresponding to available TRU of 200 t that decreases to 130 t in 2170. Then, 2 FRs from 2170 to 2230 and 2 FRs from 2230 to 2290 will be deployed. After four generations of transmutation, the amounts of Pu and MA are reduced to 40 and 30 t, respectively.

MA content of FR is as high as 15 % (Fig. 19.7), which is above the design limit of 5 % in Wakabayashi et al. [1]. In the usual design of FBRs, MA accumulation is mitigated by a supply of fresh Pu from the blanket. Moreover, high Pu content of FR burner contributes to high MA content. High MA content generally causes deteri­oration of safety parameters (beta, Doppler coefficient, void reactivity) and diffi­culty in a reprocessing and fabrication plant.

Concerns on HLW

HLW stands for high-level radioactive waste. Concern about the safety of HLW disposal is another important element for the public in deciding their choice of nuclear power along with the safety issues related to nuclear power plant operation. Former Prime Minister Koizumi changed his political stance clearly after the Fukushima nuclear accident in March 2011, from pro-nuclear to anti-nuclear, mainly on the basis of his concern about the safety of HLW disposal.

K. Yamaji (*)

Research Institute of Innovative Technology for the Earth (RITE), 9-2 Kizugawadai, Kizugawa-shi, Kyoto 619-0292, Japan e-mail: yamaji@rite. 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_24

Requirements for Long-Term Knowledge Management

As already stated, when a nuclear accident occurs there is a much larger volume of wastes and greater variety of waste types than the conventional wastes generated from the usual operation and planned decommissioning of a nuclear power unit. The wastes from a nuclear accident also contain a wider range of concentrations for the various radionuclides. Consequently, it will take a long period to characterize the wastes and to carry out the R&D and evaluation of treatment and disposal technology. It is also expected that the actual treatment and disposal of the wastes will not take place until after more than a few decades. In consideration of this waste management period, long-term knowledge management is needed. The authors propose the formation of an R&D implementation and evaluation team that will manage and retain technical knowledge. The team should involve younger staff members, and education and training of the next generation of staff should be performed by on-the-job training (OJT).

28.2 Conclusion

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 radioactivity of the water, and these methods have generated various kinds of sludge and spent adsorbents as secondary wastes. Because long-term waste management is needed, the authors examined a broad range of issues concerning how to manage these wastes 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 technology; formation of an R&D implementation and evaluation team; and long­term knowledge management are proposed.

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.