In the one-batch design in the previous section, keff drop of Pu ADSs is 14 %dk, which is too large to be compensated by burnable poison or control rods. As the first step of design improvement, a multi-batch design is introduced. Theoretically, an

Operation day

Time evolution of criticality for one-batch design

N-batch core can reduce keff drop by 1/N, that is, a drop of 14 %dk can be reduced to

2.3 %dk by six-batch design. Figure 19.4 illustrates the criticality change with an expansion for operation date of 0-1,200 days. The criticality drop for the early operation date is larger than the limit of 3 %dk, that is, 0.94 of keff at end of burn-up; the drop decreases in the equilibrium cycle. The maximum drop is 5 %dk for Pu-ADS and 6.5 %dk for Pu+U ADS, which can be compensated by control rods or burnable poison or mitigated by shorter operation in the early cycle in future improvements. The drop in the equilibrium cycle is approximately 2 %dk, which is comparable to that of the reference MA-ADS.

Volume fractions and inventories are listed in Table 19.6. Six-batch cores generally require more inventory than a one-batch core because an averaged keff during operation of multi-batch cores is higher than that of the one-batch core. Transmutation amounts of Pu — and Pu+U-ADSs are much smaller than that of the MA-ADS because the operation period is, respectively, only 50 and 100 days.

To evaluate transmutation half-life, operation and cycle efficiencies must be determined. The short operation period of 50 or 100 days implies frequent fuel exchange and low operation efficiency. There are two kinds of interval: fuel exchange and plant maintenance. We assumed that fuel exchange of a 1/6 core requires 15, 30, or 60 days for the Pu — and Pu+U-ADS and that plant maintenance including accelerator needs 60 days. Because fuel exchange for 15 days is very short, considering shutdown and startup of the ADS plant is included, tentative storage inside a core vessel should be applied for such a short interval. In the case of Pu-ADS, the 50-day operation and 15-, 30-, or 60-day interval are repeated five times, then 50-day operation and 60-day maintenance are done. In the case of Pu +U-ADS, 100-day operation and 15-, 30-, or 60-day interval are repeated two times, then 100-day operation and 60-day maintenance are done. The total operation period before a long plant maintenance of 60 days in both ADSs is 300 days.

Operation day

Table 19.6 ADS inventories for six-batch design (equilibrium core) ADS case MA (Ref., 1-batch) Pu Pu+U Volume fraction of inert matrix (%) 69.8 85.4 62.5 Core inventory at BOC (t) U 0.19 0.01 3.05 Pu 1.83 1.84 2.04 MA 2.37 0.19 0.20 Core inventory at EOC (t) U 0.18 0.01 3.01 Pu 1.79 1.80 2.00 MA 1.92 0.19 0.20 Transmutation, BOC-EOC (t) U 0.00 0.000 0.037 Pu 0.04 0.041 0.043 MA 0.45 0.000 0.000 Specific heat, h (MW/tHM) 182 393 151

Based on the foregoing assumptions, operation efficiency and cycle efficiency are determined as listed in Table 19.7, with specific heat and resulting transmuta­tion half-life. Operation efficiency multiplied by cycle efficiency of the Pu-ADS is the poorest, but the transmutation half-life is the shortest because of the high specific heat. In the present study, a 30-day interval for fuel exchange is adopted as a nominal case. The transmutation half-life of the Pu-ADS is 24.8 years in the nominal case, which is applied to scenario analysis.

Another observation is that the impact of the out-core period on cycle efficiency is significant. The out-core period is presumed considering the half-life of 242Cm of

126.8 days. If a shorter out-core period is accomplished by corresponding design of the reprocessing and fabrication plant, cycle efficiency and resulted transmutation half-life can be improved. Table 19.8 shows comparison of a 3-year and 1-year out-core period. An impact on the transmutation half-life of the Pu-ADS is a factor of around 2, and the transmutation half-life becomes as short as 13.5 years. Although 3 years of out-core period is applied as the nominal case, a shorter out-core period should be pursued in future study.

I have learned that technology for the back-end of the nuclear fuel cycle has already been well developed, but it still does not seem to be working well.

I hope that the solutions will be realized as soon as possible. Especially, the location for final disposal of nuclear wastes must be determined as soon as possible, and this is really a responsibility of the National Government to determine the location for the final disposal.

Not only Japan, but almost all countries including Germany, USA, Britain, and Russia, have not yet decided the location for final disposal, excepting Finland and Sweden. This decision must be made irrespective of whether nuclear power stations are to be continued.

A fundamental issue is to establish a raw waste inventory list with information on various characteristics, including chemical and physical form and radionuclide inventory (Table 28.1). This is the first step for pursuing further examination of treatment and disposal of these wastes. However, it is very difficult to establish a

complete list from the beginning because the volume of waste and variety of waste types is large and certain radionuclides contained in the wastes are difficult to identify and measure. Therefore, stepwise development and evaluation are impor­tant. A management system of the waste inventory should be established, and new waste information obtained by several intensive waste characterization projects should be added to the management system, where the information will be shared among concerned organizations.

The management system should include information about treatment method and source term characteristics, including waste form, volume, surface dose rate, ID number, and location of waste package, which are collected when raw wastes are treated. This system should be maintained and used until final disposal of the wastes. When all wastes are disposed of in a designated disposal facility, the inventory list will be used as the waste records of the disposal facility.

The TRIGA fuel consists of a U-metal phase and a Zr-hydride phase at high temperature in the reactor. The MA-hydrides are stable at high temperature

[8]
. The phase relationship of the U-Th-Zr hydride has been studied, considering Th as a surrogate of MA. Figure 16.1 shows the microstructure of UTh4Zr10H24: black areas are Zr hydride, gray region is ThZr2Hx, and white areas are uranium metal. The thermodynamic analysis shows that the MA-hydride consists of MA-hydride, MA-Zr-hydride, and Zr-hydride (Fig. 16.1).

Fig. 16.2 MA-hydride target pin

For the application of Hf-hydride to neutron absorber material in FBR [3], the Hf-hydride pin has been developed. The fabrication of sodium-bonded Hf-hydride pins has been demonstrated. The pins were successfully irradiated in BOR-60 for 1 year [7]. The MA-target pin was designed based on the foregoing experiences. Figure 16.2 shows a target pin that includes MA-hydride ((MA, Zr)H16) pellets. The gap of the MA hydride pellet-stainless steel cladding was filled with liquid sodium to keep the temperature of the pellets low. The results of irradiation experiments show that the sodium also reduces loss of hydrogen from the hydride pin.

Sensitivity analyses were conducted for several selected nuclides in Zircaloy-2, SUS304 stainless steel, and INCONEL alloy. Analyses in Zircaloy-4 were skipped because the sensitivity coefficients were thought to be almost the same as that in Zircaloy-2 because calculation conditions were similar. For SUS304 stainless steel, activations using the cross-section library of void ratio 0 % were evaluated because the concentrations in the case of void ratio 0 % were larger than that of void ratio 70 %.

The sensitivity coefficients of initial compositions are shown in Table 20.7. As defined in Eq. (20.1), the value shows the relative amount of variation in concen­tration of the target nuclide when the initial composition of element varies by a unit amount. Therefore, the source elements leading to the generation of target nuclides was clarified from the results. For example, Table 20.7a shows that Fe-55 is generated from both iron and nickel and that the contribution from iron is dominant. The results can also be useful in the evaluation of the error propagated from the measurement uncertainty of initial composition.

As defined in Eq. (20.2), a sensitivity coefficient of a cross section shows the relative amount of variation in the concentration of the target nuclide when the

cross section varies by a unit amount. Therefore, a positive value of this coefficient indicates that the target activation product is generated through the nuclear reaction. Thus, if a sensitivity coefficient is positive and large, the cross section of the nuclear reaction is significant for the generation of the target activation products. In the analyses, the objectives of reaction were six reactions treated in ORLIBJ40 library; the reaction of (n, y), (n, 2n), (n, a), and (n, p) yielding to nuclides of ground state and the reaction of (n, y) and (n, 2n) yielding to nuclides of meta-stable state. The summary of the results of sensitivity analyses of cross sections are shown in Table 20.8, where the sensitivity coefficients that are positive and more than 0.01 are extracted from all the results and listed in descending order. The results clarified the nuclear reaction dominating the generation of target nuclides. For example, it is thought that Fe-55 in Zircaloy-2 can be generated from the (n, y) reaction of Fe-54, the (n, a) reaction of Ni-58, and the (n, 2n) reaction of Fe-56. Table 20.8a clearly shows the (n, y) reaction of Fe-54 is dominant in the generation of Fe-55.

It was remarkable that the dominant generation pathways were clarified even for the target nuclides generated through complicated pathways. Some of the examples are shown in Fig. 20.2.

Figure 20.2a shows an example of nuclides generated with some contributing pathways. Be-10 is generated in Zircaloy-2 mainly through two pathways, the (n, p) reaction of Be-10 and the (n, a) reaction of C-13. It is not predictable which pathway is dominant from the initial composition of the material. The sensitivity

coefficients clearly showed that the (n, a) reaction of C-13 is the dominant pathway for Be-10 generation in Zircaloy-2.

Figure 20.2b shows an example of nuclides generated through long and compli­cated generation chains. The source nuclide of Cl-36 generated in SUS304 stainless steel is ambiguous because the initial composition in this analysis does not contain chlorine, which could be the dominant source element of Cl-36. The sensitivity coefficients quantitatively clarified that S-34 is the source nuclide of Cl-36 even for the long and complicated chain.

The Review Committee of SCJ consists of various experts from wide-ranging academic fields from physical science, engineering, life science, social science, and humanities. The proposal concerning awareness of the limits of scientific and technical abilities was formed through interdisciplinary discussions among the experts. Some readers of the SCJ report seem to have felt uneasiness with this proposal because this proposal apparently cast a scientific doubt on the feasibility of the geological disposal of HLW. To the author’s understanding, this proposal is a rather general statement that there is no perfect scientific evidence to support the safety of HLW disposal for more than 10,000 years.

Having heard the discussions related to this proposal, the author recognized there are many different academic approaches depending on the field of science. For example, natural scientists seek truths in natural phenomena, whereas engineers try to make things and/or systems that are valuable and acceptable for human society. HLW issues are related not only to various fields of science, but also to value systems shared by society. Here again, the author was convinced that we need to reflect more deeply on the relationship between science and society.

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.

Didier Haas, M. Hugon, and M. Verwerft

Abstract Since the early 1970s, studies and experimental projects have been undertaken in Europe to examine the potential of thorium-based fuels in a variety of reactor types. The first trials were mainly devoted to the use of thorium in high — temperature reactors. These projects can be seen as scientific successes but were not pursued on a commercial basis because of the priority given in Europe to the development of light water reactors. Later on, thorium oxide was considered as a potential matrix for burning plutonium (possibly also minor actinides), and several core design studies, as well as experiments, were undertaken. The most recent such concern the BR2 and HFR Material Test Reactor (MTR) irradiations in Belgium and in the Netherlands, respectively, as well as the KWO PWR in Obrigheim in Germany, in which thorium-plutonium oxide fuel (Th-MOX) was successfully irradiated up to 38 GWd/tHM. The results of these experiments have shown that Th-MOX behaves in a comparable way as conventional uranium-plutonium oxide fuel (U-MOX). More work is still needed before Th-MOX will reach sufficient maturity to implement it on a large scale in power reactors, but all currently available results indicate that licensing Th-MOX for LWRs should be feasible. Finally, European research projects are still devoted to the study of thorium salts in molten salt reactors, a design that incorporates on-line reprocessing and needs no specific thorium fabrication, adding therefore the benefits of thorium without its main challenges.

Keywords Fuel • HTR • LWR • MSR • Plutonium • Thorium

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

D. Haas (*)

Consultant, 101 rue de la Station, 1457 Walhain, Belgium e-mail: Didier. haas@hotmail. be

M. Hugon

European Commission, Brussels, 1049 Brussels, Belgium M. Verwerft

SCK*CEN, Boeretang 200 B-2400 Mol, Belgium © The Author(s) 2015

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_18

18.1 Introduction

Natural thorium (Th) has only one isotope, 232Th, which is fertile. In a thermal reactor, Th can absorb neutrons and, following nuclear reactions, produces 233U, which is fissile. Under optimized breeding conditions, a sustainable Th-233U cycle can be reached, but the thorium cycle needs a seed or driver fuel, which can be based on 235U or on Pu.

233U as a fissile nuclide features high neutron production in a thermal and epithermal neutron spectrum. This ability offers improved neutron economy for reactors fueled with U rather than U or 9Pu, particularly at thermal energies in light water reactors (LWRs). In theory, breeding (formation of fissile nuclides) is achievable at thermal energies with a Th/233U fuel, which is not the case with U-MOX fuel. However, even though breeding can be demonstrated at an experi­mental level, optimal breeding is not achieved in the current fleet of LWRs. In today’s context, U-MOX fuels are not reprocessed, and here Th-MOX offers perhaps its best advantage over U-MOX. The excellent chemical stability of the thorium oxide matrix makes it an excellent candidate for direct disposal, and thus also for once-through fuels allowing burning excess Pu without production of higher actinides. Another alternative would be to use Th-MOX fuels in LWRs as a means to initiate the breeding of 233U for future use in other reactor types, as an option to save natural U and to further improve the U-Pu fuel cycle.

In addition to the LWR/FR scenario, two reactor types have been considered for a breeding Th fuel cycle in the future: high-temperature reactors (HTRs) and molten salt reactors (MSRs). HTRs represent the fastest route to implement a closed breeding Th fuel cycle. The technology exists conceptually but needs to be devel­oped before commercialization (which is pending). Also, supporting technologies associated with fuel manufacturing, reprocessing, transport, waste management, and final disposal need to be developed. MSRs represent a longer-term develop­ment option for Th fuel cycles. In MSRs loaded with Th-based fuels, breeding may be achieved over a wide range of neutron energies. On-line reprocessing is an important feature of MSRs, which enables continuous re-use of the nuclear fuel by extracting the fission products.

The potential development of a closed Th fuel cycle faces some obstacles. Reprocessing is one of these, as Th oxide is more stable than U oxide. In contrast to the Purex process, which has been industrially operational in the U-Pu fuel cycle for more than 30 years, the Thorex process, which has been investigated for many years in laboratories, faces some difficulties: it requires stronger acids (and there­fore more advanced corrosion-free materials for process vessels) and longer disso­lution times. Remote-controlled fuel manufacturing represents another challenge as Th-based fuels have high-energy gamma radiation from the presence of 232U after irradiation, which requires remote fabrication and handling in heavily shielded facilities. Thus, this fuel fabrication, transport, and reprocessing are more complex than the present practice for U oxide fuel, for instance.

This presentation summarizes the history and status of the main European research programs (cordis. europa. eu) with Th use. These programs concerned HTRs, LWRs, and MSRs. Emphasis is given here on the latest two developments.

22.2.1 Modification of STACY

To implement the new criticality control measures for fuel debris, the Japan Atomic Energy Agency (JAEA) has been carrying a project to modify the Static Experiment Critical Facility (STACY) to pursue critical experiments on fuel debris [6]. STACY, a facility using solution fuel (low-enriched uranyl nitrate), is to be converted into a thermal critical assembly using fuel rods and a light water moderator.

In the modified STACY, the core configuration consists of fuel rods loaded in the core tank (up to 900 rods) and light water fed as moderator. Because the maximum thermal power is only 200 W, fuel burn-up is negligibly small and cooling water is unnecessary. The reactivity of the core is controlled not with control rods but by water level, and with safety plates (cadmium) in the case of emergency shutdown, similar to the present STACY. The fuel rods contain 5 wt.%-enriched UO2 pellets and have zircaloy cladding. A soluble neutron poison (boron) can be added to the light water moderator. Major core specifications and a schematic diagram of the modified STACY are shown in Table 22.1 and Fig. 22.1, respectively.

We carried out batch sorption experiments using 63 Japanese rice paddy soil samples to clarify the transfer pathways of 14C in rice paddy fields. The soil samples were collected throughout Japan and taken to our laboratory where they were air dried and sieved (<2 mm). These sieved soils were mixed with a [1,2-14C] sodium acetate solution at the ratio of soil : solution = 0.5 g: 5 ml, and the flooded soil samples were incubated at 25 °C for 7 days [2]. During the incubation period, the 14C atoms of the sodium acetate were partitioned into solid, liquid, and gas phases. Each partitioning ratio is shown in Fig. 26.1. Approximately 63 % of the total 14C on average was released into the air as gaseous compounds. Partitioning ratios into solid and liquid phases were 34 % and 3 %, respectively. These results suggest that gasification is an important pathway in the environmental transfer of 14C in Japanese rice paddy fields.

When 14C is released into the air, 14C-bearing gases must pass through the soil solution. Because soil solution pH affects chemical reactions such as hydrolysis and degassing of CO2, chemical forms of 14C-bearing gases may change in the soil solution. We, therefore, investigated relationships between pH and partitioning ratios of 14C into the liquid phase at day 7 of incubation (Fig. 26.2). The partitioning ratio increased with increasing in pH, and a significant correlation (r = 0.7) was found. These data fit well with the solubility curve of total carbonic acid in water, which refers to the sum of dissolved carbon dioxide and the carbonic acid. This observation suggested that the dominant chemical species of 14C in gas forms was carbon dioxide. To confirm the effect of pH on the partitioning of 14C into the liquid phase, a soil sample was suspended in MES [2-(N-morpholino)ethanesulfonic acid] buffers with the initial pH value adjusted to 5.5, 6.5, and 7.5 (Fig. 26.3). A control

Fig. 26.1 Box plots for each partitioning ratio of 14C into solid, liquid, and gas phases

sample was prepared consisting of the soil and deionized water (pH unadjusted). The partitioning ratio also increased with increasing pH, suggesting that the partitioning ratio of 14C into the liquid phase depended on the pH of the soil solution.

Soil-soil solution distribution coefficient (Kd) is a commonly used parameter to evaluate behaviors of radionuclides in the environment. In our study, the Kd values were calculated from activities of the 14C in the solid and liquid phases at the end of incubation, and the obtained Kd value was 139 ± 77 ml g-1 on average. Negatively charged anions generally have low Kd values because of simple electrostatic interaction. Our value, however, was higher than expected from the chemical form of 14CH314COO-. For example, Kaneko et al. [1] obtained the Kd value of

9.5 ml g-1 for the sorption test of acetic acid using cement materials. The reason for our high Kd value is explained next.

Figure 19.5 illustrates a result of the LWR-OT scenario where time evolution of electricity generation, Pu inventory, and MA inventory are shown. The peak of 50 GWe appears in 2010 and decreases because of the Fukushima accident and closure after 40-year operations. All LWRs will be shut down in 2055. The Rokkasho reprocessing plant (RRP) will not be operated, but 7,100 tHM spent fuel has been reprocessed, mainly overseas. The year of reprocessing is not clear

but assumed to be in the 1990s. A small amount of MOX fuel from this reprocessing will be utilized in LWRs.

Pu inventory mainly exists in UO2-SF. “Pu” in the figure is not “separated” Pu, but Pu in MOX fresh fuel in this scenario. The total of plutonium is 350 t that is gradually disposed of to a repository from 2043 until 2105. The trend of MA inventory is almost the same, but it continues to increase after 2040 because 241Pu becomes 241Am with a half-life of 14.35 years.