The last option is, in fact, being currently applied. The defense line consists of xenon gas monitoring and the injection of borated water. The monitoring sensitivity is not sufficient to measure subcriticality but can detect the event beyond the occurrence of critical condition before severe consequences result. The borated water on standby will be injected when the monitoring detects the criticality. A study is under way to improve the monitoring sensitivity to make the detection and intervention quicker and to reduce the risk of this option.

A much bolder idea is also being brought up, which is to consider such quick detection and intervention as a regular reactivity control. A small-scale, controlled chain reaction is permissible in the concept, and the resumption of fuel debris retrieval is allowed after suppressing the criticality. To realize this kind of critical­ity control, its risk must be fully understood.

The results of this study suggest that when dealing in class with social issues such as the disposal of high-level radioactive waste that require a certain level of basic scientific knowledge, debate and other kinds of active learning may be more effective than lecture-style, noninteractive pedagogical methods. When implementing active learning, however, time and cost and many other factors need to be considered. For the courses described in this chapter, the author was

able to gain the cooperation of NUMO, which made it possible for her to integrate into the course lectures by experts and visits by students to relevant facilities. But when such outside help cannot be obtained, alternatives must be considered.

To take debate on social issues beyond the walls of the university classroom, a DVD of the 2013 students’ debate competition may be a valuable resource that could be shown to the general public. In addition, there is a need for a more conducive learning environment for the debating, which could perhaps be achieved by creating scenarios for model debates and organizing workshops.

This course also pointed to the need for consumer education. It is important to not only think about the short-term consequences of the things we use and their disposal, but also to consider from a global perspective how our consumption will affect future generations. In debate for education, it is important to tackle a range of social issues from multiple points of view. In debating the issue of radioactive waste, in particular, it must be conveyed to students that the issue is not one that is solved just by shutting down nuclear reactors; whether or not we use nuclear energy in the future, the waste has already been produced, a by-product of our consumption.

This chapter has focused on courses aimed at undergraduate students. If possi­ble, in the future, the author would like to run similar debating courses with middle — and high-school students and with adults, and then assess, as was done in this study, the impact of debate as an educational activity.

19.3.1 Reference ADS (MA-ADS)

An ADS core dedicated for MA transmutation (MA-ADS) [8] bases the present design for Pu transmutation. As listed in Table 19.1, the MA-ADS is a medium-size core loaded with nitride fuel cooled by lead-bismuth eutectic (LBE). Nitride fuel is diluted by zirconium-nitride with a weight fraction of around 50 % (volume fraction around 65 %) that can be changed so that criticality becomes an appropriate level (keff = 0.97). The ADS is operated for 600 effective full-power days (EFPDs) without any fuel reloading while the interval for maintenance of an accelerator can occur. In other words, the ADS is a one-batch core that implies large reactivity drop after depletion in the case of Pu transmutation. Because control rods for criticality are not equipped in the ADS, the drop must be supplemented by

increasing proton beam current. Fortunately, MAs are fertile material that becomes fissile after capturing one neutron, and the drop is very small, as shown in the following section. Table 19.3 provides weight composition of MA fed to MA-ADS. Figure 19.2 illustrates the R-Z model for calculation.

According to the world energy outlook of IEA, the total electric power generation will be increased as shown in Fig. 23.3. The electric power generated by renewable energy is predicted as shown in Fig. 23.4. The electric power generated by renew­able energy other than water power will increase very slowly, from about 4 % in 2010 to only 15 % in 2035, whereas the electric power generated by nuclear energy will be kept almost constant from 13 % in 2010 to 12 % in 2035.

We must try to increase renewable energy more as quickly as possible.

In this respect, I commend Germany, which has strived to increase the devel­opment of renewable energy (Fig. 23.5) after 2000. In 2010, electricity generated by renewable energy reached 103.5 billion kWh. Deducting that generated by water power, we have 82.9 billion kWh. Total electric power generation in Japan was 976.2 billion kWh in 2010; that is, electric power generated by renewable energy other than water in Germany in 2010 was only 8.5 % of the total electric power generation in Japan in the same year. The electric power generated by nuclear energy in Japan was 300.4 billion kWh in 2010. Therefore, electric power generated in Germany by renewable energy sources other than water in 2010 is only 28 % of this amount. Germany has striven so much in these 10 years from 2000 to 2010, and the average price of electricity per house has doubled; that is, Germany has invested a large budget. It takes many years to increase renewable energy, and the result is still not satisfactory. Even if Japan tries as much as Germany, it will takes at least 30 years to replace nuclear energy by renewable energy. Meanwhile, Japan must depend on fossil fuel, which increases CO2 emissions into the atmosphere. To import fossil fuel, the deficit in foreign trade of Japan, which is now already more than 4 trillion yen (about $40 billion), will continue as the result of the decrease in nuclear energy.

When we stop all nuclear power stations in Japan, renewable energy must be increased, not only to replace nuclear energy but also the energy produced by fossil fuel. Is this really possible in the near future? It is time for us to deliberate upon the future of energy in Japan to guarantee energy security, to avoid global warming, and to stabilize the economy of Japan.

Fig. 23.3 World energy outlook of the International Energy Agency (IEA)

Share of renewables in world electricity production Fig. 23.4 Electric power generated by renewable energy (prediction by IEA)

27.4.1 Separation Using Anion-SR

Good recovery of I_ (as 129I~) from the Anion-SR was observed in the 3 M NaOH solution, but only minor amounts of IO3~ (as 127IO3~) were recovered (Table 27.1). The percent recoveries did not appear to be dependent on the reductant because good recoveries of I_ and poor recoveries of IO3~ were observed regardless of the presence or absence of NaHSO3. This result supports the expectation that 129I~ is extracted and ‘IO3 is not extracted by the Anion-SR, as ‘IO3 was not reduced to 127I~ by NaHSO3 in 3 M NaOH solution, and the isotopic exchange reaction between 127IO3~ and 129I~ and reduction from 127IO3~ to 127I~ were negligible at least for 1 day. It is reported that 129I in seawater offshore of Fukushima is mainly 9I regardless of the presence of large amounts of natural ‘IO3 [8].

In the diluted HCl solution at pH 2, 129I and 127I were recovered in the presence of the reductant and were not recovered without the reductant (Table 27.1). Both the isotopes behaved similarly in this case despite the difference in the initial chemical species, I and IO3 . In contrast, the experiments using 1 ^g I in the

diluted HCl solution at pH 2 without the reductant and standing, higher recovery of 127I~, 72 ± 6 %, was obtained. Although the influence of the I_ amount on recovery cannot be wholly denied, it is possible that a considerable amount of 129I~ would
be changed to another chemical species in the diluted HCl solution in a day. In addition, the changed chemical species and IO3~ were reduced to I_ by the reductant in this experimental condition. Consequently, inorganic iodine species were analyzed at pH 2 with the reductant, and only I_ was analyzed in 3 M NaOH. It is possibly to apply speciation methods to analyze I_ and IO3~, depending on the solution conditions.

The core performance of the developed uranium-free core was evaluated as shown in Table 15.4. The Zr content in the fuel alloy was determined to maintain criticality during the operation cycle under the conditions of the upper limit of the melting point, 1,200 °C. According to the results, the uranium-free TRU metallic core is viable in terms of core performance, safety performance, fuel fabrication, and TRU burner.

The Doppler coefficient is similar to that of the conventional metallic fuel fast reactor cores, and the burn-up reactivity swing is considered to be controllable by conventional control rods and fixed absorbers. Moreover, core sodium void reac­tivity including the upper plenum region is negative because of neutron leakage at the upper plenum region and neutron spectrum moderation from the presence of BeO during sodium voiding. Although the restriction for sodium void reactivity

was not assumed for the core in this study, low sodium void reactivity is a significant factor for sodium-cooled fast reactors.

Furthermore, the developed core design has the potential to achieve passive safety features against unprotected events such as unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) similar to a conventional metallic fuel core because the basic core safety parameters, that is, average and peak linear heat rates for lower fuel temperatures, the enhanced Doppler coefficient, and low sodium void coefficient (negative sodium coefficient in whole core), were maintained within the similar ranges of a conventional metallic fuel core design [20].

Feasibility in the light of decay heat is also confirmed to be practicable, as the decay heat of the fresh fuel material is 32 W/kgHM, which is less than 10 % of that of the minor actinide (MA)-only fuel. Also, the decay heat of the fresh fuel subassembly is approximately 240 W. Taking advantage of some cooling scheme such as air flow, this fuel can be fabricated as a fuel pin bundle [21].

Moreover, the results also shows the profitability of the uranium-free TRU metallic fuel fast reactor itself, because a 1-year operation of this 300 MWe TRU-burning fast reactor burns 260 kg TRU, corresponding to the amount pro­duced by a 1.2 GWe-year operation of a conventional LWR.

For all these reasons, the TRU-burning fast reactor using uranium-free TRU metallic fuel is considered to be feasible. Further study such as reduction of burn-up reactivity swing and trade-off of various countermeasures considering economic aspect helps improve and optimize the core design in the next phase.

15.3 Conclusions

A TRU transmutation system associated with the uranium-free metallic fuel fast reactor is a practical way to burn TRU with sustainability, fewer R&D needs, and a simple system, because it can be used as both a TRU burner and a power supply plant. Employment of pyro-processing for recycling reduces the burden of R&D requirements, and introduction of a conventional fuel fabrication method and pyro — processing allows less complex facilities.

In this study, two main issues related to the uranium-free core were investigated and discussed to clarify the feasibility of a TRU-burning fast reactor cycle using such a core: Doppler coefficient for reactor safety, and burn-up reactivity swing for acceptable reactor operating cycle length.

The results show that the uranium-free fast TRU fast reactor core is viable because those issues can be solved by TRU-Zr alloy fuel, BeO neutron moderator, and reduced core height. Thanks to the BeO pins that function not only as a neutron moderator but also as a diluent material, the 35 %Zr alloy fuel can be fabricated without Am vaporization because its melting point is maintained below 1,200 °C, the temperature that causes Am vaporization during injection casting fuel fabrica­tion. Moreover, the decay heat of the fresh fuel is considered to be an acceptable level for the fuel fabrication. Also, a 1-year operation of this 300 MWe core burns the TRU that is produced by 1.2 GWe-year operation of a conventional LWR.

In conclusion, the prospect of a TRU-burning fast reactor cycle using uranium- free metallic fuel was confirmed. Further study, not only to improve core perfor­mances but also to develop a recycling process associated with this uranium-free system, which is currently under way, promotes realization of the system.

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.

20.3.1 Analyses Conditions

As stated in Sect. 20.1, activations of in-core structure materials, such as cladding tubes, end plugs, and spacers of fuel assemblies and channel boxes, were investi­gated in this study. The materials of the in-core structures of PWR and BWR are shown in Table 20.1. The compositions of Zircaloy-2, Zircaloy-4, SUS304 stainless steel, and INCONEL alloy 718 are shown in Table 20.2. In Table 20.2, the average value of the upper and lower limits of the standard specification was applied to the calculation condition for additive elements and the upper limit was applied for impurity elements. The effect of impurity elements that are not specified in the standard are investigated in Sect. 20.3.4.

Typical conditions of BWR were assumed for the cross-section libraries and the irradiation condition, because the difference between the conditions of PWR and BWR is not so significant for the purpose of this study, which is clarifying the dominant generation pathways of activation products.

The cross-section libraries used in these analyses (Table 20.3) were chosen to correspond to the condition of the void ratio in the axial direction. A library made with an average void ratio (40 %) was applied to cladding tubes, spacers, and channel boxes for which the void ratio varies from 0 % to 70 %.

A BWR typical irradiation history consists of four cycles of irradiation of about 377 days with constant flux and 90 days of cooling time in the intervals of irradiation (Fig. 20.1). Considering the period for processing of radioactive wastes,

10 years of cooling time after irradiation was assumed in these analyses. The flux intensities at the center, top, and bottom in the axial direction are shown in Table 20.4. The flux intensity at the center corresponds to the average power in typical BWR fuel assemblies. The flux intensities at the top and bottom were

determined to be 5 % of that at the center, based on flux distribution evaluated by the one-dimensional neutron diffusion calculation.

Proposals of the SCJ report, particularly, the concepts of temporal safe storage and management of total amount, triggered many discussions widely concerning the issue of HLW disposal.

The temporal safe storage is characterized by securing a moratorium period of several dozen or several hundred years to establish appropriate handling measures for the problem. It provides the advantages of using this period to refine techno­logical developments and scientific knowledge, guaranteeing the possibility of creating handling measures that target a longer period; for example, improvement of the durability of containers, development of nuclear transmutation technology to reduce volume and toxicity of HLW, and research related to the stability of geological layers.

In addition, the temporal safe storage makes it possible to keep various options for future generations to choose for final disposal of HLW.

The concept of safe storage, however, still has a wide range of uncertainties in technical specifications; for example, duration of storage, location character­istics such as on ground or underground, and number of storage facilities. The concept ranges from currently available interim storage of spent fuel to retriev­able geological disposal. In fact, the response of Japan Atomic Energy Com­mission mentioned retrievable geological disposal in the context of temporal safe storage.

SCJ had set up a Follow-up Committee as an extension of the Review Commit­tee in August 2013 to clarify the concept of the temporal safe storage.

In sensitivity and uncertainty analysis, multi-group sensitivities are usually used, but there is no theoretical basis for the effect of number of energy groups to sensitivities. Here we derive a relationship between sensitivities calculated with different numbers of energy groups by considering the case where multi-groups are collapsed to a few groups. The sensitivity of core parameter R to the microscopic cross section of nuclide i and reaction j in group g in multi-groups is denoted by S and is defined by

The sensitivity of R to microscopic cross section in few groups is given by

dR/R

d*fi/*5

Cross sections from few groups are calculated from a multi-group cross section by using neutron flux ф8 in group g:

where the summation about g is performed over energy groups g included in few groups G. Let us consider the case where multi-group cross sections change as follows:

With the cross-section change, the neutron flux also changes:

Ф8 —— ф8 + $ф8

Therefore we obtain

Here we apply the narrow resonance approximation to express the flux perturbation caused by cross-section change.

where C is a constant, Ni is number density of nuclide i, eg is microscopic total cross section of nuclide i, and o0 is background cross section. When using only the j reaction cross section of nuclide i, og, the flux perturbation is expressed by

where o0 — o0 + — і*] in the first-order approximation. Introducing the preced-

j—j

ing equation to Eq. (17.25) leads to

We change the multi-group cross sections eg at constant rate a (for example, 1 %) within few groups G:

In this case, the few-groups cross-section change is expressed by the multi-group sensitivity as follows

SG=£ Sg

g є G

We use this relationship to choose energy groups N (G = 1-N) such that

SG ^2 sg

g є G

As an example, we calculated keff sensitivities in 7, 33, and 70 energy groups, and compared sensitivities. In 7 groups, the sensitivities to 235U capture cross section are different from the corresponding integrated sensitivities calculated from 70 groups by 10-20 % above 100 eV. However, in 33 groups, the sensitivities are different from the 70 groups result by at most 5 %. This result convinced us that calculations of sensitivities for 33 groups or 70 groups are sufficient.

The risk study is necessary regardless of which option is chosen because the subcritical condition is not secured at present. Even though the fuel debris will not be touched for a while, the temperature of the fuel debris may drop gradually in time, which slowly increases reactivity. The risk of “low probability and high consequence events” must be also evaluated. An aftershock of large magnitude may change the fuel debris geometry greatly. The extreme event would be the fall of fuel debris in the PV onto the other in CV.

The fuel debris retrieval must be assessed carefully, of course, if it is conducted under nonborated water. The first step of the risk analysis is to understand the actual conditions of fuel debris. Exhaustive observation of the fuel debris should be conducted as early as possible, which enables us to complete the maps described in the previous sections.

According to each option, engineering work should be performed in parallel to establish design requirements. For the prevention of criticality by borated water, its required lowest concentration must be established. For the prevention of criticality by monitoring, requirements of sensitivity and time response of the monitoring and time response of an intervention measure must be clarified. For the prevention of severe consequences, an allowable limit of fission number must first be set. Then, the time response of detection and intervention must be defined to regulate fission numbers of supposed criticality events within the limit.

21.5 Conclusions

In 1FNPS, fuel debris conditions in the three damaged reactors are still unknown and uncertain. The water issue also affects criticality control, as the coolant water is not borated. Although fortunately no sign of criticality has yet been seen, the subcritical condition is not secured. There are options of principles to pursue a certain critical control of the fuel debris: prevention of criticality by poison, by dry process, or by monitoring, and prevention of the severe consequences resulting from criticality. Engineering research and development is to be conducted regard­ing any of these options.

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