The schematic diagram of the LBE test loop is illustrated in Fig. 11.3. The test loop consists of a test section, a gas injector, an electromagnetic pump, a flow meter, and a drain tank. The test section is a stainless steel pipe with an inner diameter of 50 mm and a length of 2,000 mm. The working fluids are molten LBE and nitrogen gas. The flow rate of LBE was measured by the magnetic flow meter. Nitrogen gas was injected into LBE flow by the gas injector, which consists of 101 stainless steel needle tubes 0.58 mm in inner diameter. The gas flow rate was controlled by a mass flow controller. The operating temperature of this loop was maintained at 200 ° C and the heating power was controlled by a temperature controller unit. The flow rate, differential pressure, temperature, and liquid level were monitored by a data acquisition unit connected to a PC. In the experiments, the superficial gas and liquid velocities were varied. Three four-sensor probes or electromagnetic probes were installed at three different axial positions (z/D = 3.2, 17.6, and 32.4) of the test section to investigate the axial development of two-phase flow structure. In addi­tion, these probes were traversed at 12 radial points to obtain the radial profiles.

The accelerator is the driver of MYRRHA because it provides the high-energy protons that are used in the spallation target to create neutrons, which in turn feed the core. In the current design of MYRRHA, the machine must be able to provide a proton beam with energy of 600 MeV and an average beam current of 3.2 mA. The beam is delivered to the core in continuous wave (CW) mode. Once per second, the beam is shut off for 200 qs so that accurate on-line measurements and monitoring of the subcriticality of the reactor can take place. The beam is delivered to the core from above through a beam window.

Accelerator availability is a crucial issue for the operation of the ADS. A high availability is expressed by a long mean time between failure (MTBF), which is commonly obtained by a combination of overdesign and redundancy. In addition to these two strategies, fault tolerance must be implemented to obtain the required MTBF. Fault tolerance will allow the accelerator to recover the beam within a beam trip duration tolerance after failure of a single component. In the MYRRHA case, the beam trip duration tolerance is 3 s. Within an operational period of MYRRHA, the number of allowed beam trips exceeding 3 s must remain under 10. Shorter beam trips are allowed without limitations. The combination of redundancy and fault tolerance should allow obtaining a MTBF value in excess of 250 h.

At present, proton accelerators with megawatt-level beam power in CW mode only exist in two basic concepts: sector-focused cyclotrons and linear accelerators (linacs). Cyclotrons are an attractive option with respect to construction costs, but they do not have any modularity, which means that a fault tolerance scheme cannot be implemented. Also, an upgrade of its beam energy and intensity for industrial application presently is not a realistic option. A linear accelerator, especially if made superconducting, has the potential for implementing a fault tolerance scheme and offers a high modularity, resulting in the possibility to recover the beam within a short time and increasing the beam energy and intensity toward industrial application of ADS technology.

1.3.1 Nuclear Transmutation of Cs with Laser Compton Scattering

Figure 1.5 shows the dependence of the reduction of 1 g 137Cs on the photon flux Ny = 1012, 1018, 1019,1020/s, which is calculated with this setup (Fig. 1.4). The number of 137Cs is effectively reduced with photon flux over 1018/s, that is, the number of 137Cs is reduced by 10 % for 24 h irradiation. Figure 1.6 shows the number of Cs isotopes when 1 g 137Cs is irradiated with photon flux 2 x 1012/s with the same setup. From this figure, we can see that the reduction rate of 137Cs by the transmutation, which is nearly equal to the generation rate of 136Cs, is two orders of magnitude smaller than the natural decay rate of 1 g 137Cs. Thus, the transmutation of 137Cs is not effective with photon flux 2 x 1012/s, which is maximum with present accelerator systems.

9.3.2.1 Static Experiments

the profile of neutron flux for the 232Th capture reactions was estimated through the horizontal measurement of 115In(n, y)116mIn reaction rate distribution, as well as described in Sect. 9.3.1.1. The wire was set in an aluminum guide tube, from the tungsten target to the center of the fuel region [from the position of (13, 14 — A0) to that of (13, 14 -1); Fig. 9.3], at the middle height of the fuel assembly. The absolute values of the measured reaction rates (Fig. 9.7) revealed differently the variation of

reaction rates attributed to varying the neutron spectrum in the core, when the spallation neutrons generated by 100 MeV protons were injected into the core. The moderating effect of the high-energy neutrons in some cores (Th-PE, Th-HEU-PE, and NU-PE: keff = 0.00613, 0.58754, and 0.50867, respectively) was observed around the boundary between the core and polyethylene regions. The 115In(n, y)116mIn reaction rates in the NU-PE core were higher than in other cores, demon­strating, that the reaction rates of 238U in the NU-PE core were larger than those of 232Th in the thorium cores with the use of 100 MeV protons. Additionally, the effect of the neutron spectrum on the reaction rates was observed with 100 MeV protons by comparing the measured results of reaction rates shown in Fig. 9.5. Thus, an expected physical effect was indeed observed as a result of the neutron spectrum change obtained by varying the moderator materials in the fuel assembly. Addi­tionally, the accuracy [5] of experimental and numerical analyses was compared successfully with the ratio (C/E) of calculations and experiments around the relative difference of 10 %, through the subcritical parameter of neutron multiplication M.

In this chapter, applicability of the self-indication method to identify and quantify nuclides in a BWR-MOX pellet is evaluated. The burnup of the MOX pellet is 0 GWd/t, 20 GWd/t, and 30 GWd/t. A plutonium vector in the fresh MOX pellet is employed as the OECD/NEA BWR MOX benchmark (Pu4) (235U, 0.2 w/o; total Pu, 6.71 w/o; 238Pu, 2.2 %; 239Pu, 46.2 %; 240Pu, 29.4 %; 241Pu, 8.8 %) [1]. The burn-up calculations of the BWR-MOX pellet are carried out by using deterministic neutronics code SARC 2006 [2] with JENDL-4.0 [3]. The numerical validations are performed by using the MVP2.0 [4] with the JENDL-4.0. The MVP2.0 is a

continuous-energy Monte Carlo code developed by the Japan Atomic Energy Agency.

The 12-m measurement line in the KUR-LINAC is simulated as a calculational geometry shown in Fig. 4.1. Figure 4.2 shows a neutron spectrum in a tantalum target that is a neutron source of the KUR-LINAC. The spectrum is calculated by MVP2.0. Using the spectrum as the surface source, the validation is carried out.

Interfacial area concentration measured by the four sensor probes was compared with existing correlations (Fig. 11.5). The vertical axis shows the estimation error between the measured and calculated interfacial area concentration. All the corre­lation overestimates the interfacial area concentration by 50-90 %, which might be caused by the differences in bubble size and shape. Most of the correlations were formulated with air-water two-phase flow data for a bubbly flow regime. However, the bubble shape in a liquid metal two-phase flow might be strongly distorted by the momentum exchange at the gas-liquid interface. Thus, a more appropriate expres­sion of the interfacial area concentration for liquid metal two-phase flow should be developed based on the experimental database.

v

5-І

О

s-

s­

w

d

о

Void fraction, a [-]

Because MYRRHA is a pool-type ADS, the reactor vessel houses all the primary systems. In previous designs of MYRRHA, an outer vessel served as secondary containment in case the reactor vessel leaks or breaks. In the current design, the reactor pit fulfills this function, improving the capabilities of the reactor vault air cooling system. The vessel is closed by the reactor cover, which supports all the in-vessel components. A diaphragm, inside the vessel, acts to separate the hot and cold LBE plenums; it supports the in-vessel fuel storage (IVFS) and provides a pressure separation. The core is held in place by the core support structure consisting of a core barrel and a core support plate. Figure 7.2 shows vertical cut sections of the MYRRHA reactor showing its main internal components.

At the present state of the design, the reactor core (Fig. 7.3) consists of mixed oxide (MOX) fuel pins, typical for fast reactors. In subcritical mode, the central hexagon houses a window beam tube-type spallation target. Thirty-seven positions can be occupied by in-pile test sections (IPS) or by the spallation target (the central one of the core in subcritical configuration) or by control and shutdown rods (in the core critical configuration). This design gives a large flexibility in the choice of the more suitable position (neutron flux) for each experiment.

The requested high fast flux intensity has been obtained by optimizing the core configuration geometry (fuel rod diameter and pitch) and maximizing the power density. We will be using, for the first core loadings, 15-15Ti stabilized stainless steel as cladding material instead of T91 ferritic-martensitic steel that will be qualified progressively further on during MYRRHA operation for a later use. The use of lead-bismuth eutectic (LBE) as coolant permits lowering the core inlet operating temperature (down to 270 °C), decreasing the risk of corrosion and allowing increasing the core ДТ. This design, together with the adoption of reliable and passive shutdown systems, will allow meeting the high fast flux intensity target.

In subcritical mode, the accelerator (as described in the previous section) is the driver of the system. It provides the high-energy protons that are used in the spallation target to create neutrons which in their turn feed the subcritical core. In subcritical mode the spallation target assembly, located in the central position of the

F. In-vessel Fuel Handling Machine

H. Above Core Structure

E. Primary Pump

Fig. 7.2 Section of the MYRRHA-FASTEF reactor core, brings the proton beam via the beam tube into the central core region. The spallation heat deposit is dissipated to the reactor primary circuit. The spallation module guarantees the barrier between the reactor LBE and the reactor hall and ensures optimal conditions for the spallation reaction. The spallation module assembly is conceived as an IPS and is easily removable or replaceable.

The primary, secondary, and tertiary cooling systems have been designed to evacuate a maximum thermal core power of 110 MW. The 10 MW more than the nominal core power account for the power deposited by the protons, for the power of in-vessel fuel, and for the power deposited in the structures by y-heating. The average coolant temperature increase in the core in nominal conditions is 140 ° C with a coolant velocity of 2 m/s. The primary cooling system consists of two pumps and four primary heat exchangers (PHX).

The interference of the core with the proton beam, the fact that the room located directly above the core will be occupied by much instrumentation and IPS pene­trations, and core compactness result in insufficient space for fuel handling to (un)load the core from above. Since the very first design of MYRRHA, fuel handling has been performed from underneath the core. Fuel assemblies are kept by buoyancy under the core support plate.

Ш 57 FA

7 (central) IPS 6 CR (buoyancy)

О 3 SR (gravity)

36 "inner" Dummy (LBE) 42 "outer" Dummy (YZrO) 1S1 S/As

Two fuel-handling machines are used, located at opposite sides of the core (Fig. 7.4). Each machine covers one side of the core. The use of two machines provides sufficient range to cover the necessary fuel storage positions without the need of an increase for the reactor vessel when only one fuel-handling machine is used. Each machine is based on the well-known fast reactor technology of the ‘rotating plug’ concept using SCARA (Selective Compliant Assembly Robot Arm) robots. To extract or insert the fuel assemblies, the robot arm can move up or down for about 2 m. A gripper and guide arm is used to handle the FAs: the gripper locks the FA, and the guide has two functions, namely to hold the FA in the vertical orientation and to ensure neighboring FAs are not disturbed when a FA is extracted from the core. An ultrasonic (US) sensor is used to uniquely identify the FAs.

The in-vessel fuel-handling machine will also perform in-vessel inspection and recovery of an unconstrained FA. Incremental single-point scanning of the dia­phragm can be performed by an US sensor mounted at the gripper of the IVFHM. The baffle under the diaphragm is crucial for the strategy as it limits the work area where inspection and recovery are needed. It eliminates also the need of additional recovery and inspection manipulators, prevents items from migrating into the space between the diaphragm and the reactor cover, and permits side scanning.

Table 1.1 summarizes the generation rate ofA-1X from of 1 g of a fission product AX with photon flux 2 x 1012/s also for 129I, 135Cs, and 90Sr in addition to 137Cs. The reduction rate, which is nearly equal to the generation rate of A-1X, does not depend significantly on nuclides because the properties of GDR are smooth func­tions of the mass number. From this table, we can see that the reduction rate for the transmutation of 137Cs can be similar to that of other medium nuclides.

1.2

Conclusion

In this work, the effectiveness of transmutation with laser Compton scattering for reducing fission products was quantitatively investigated. The transmutation of 137Cs is effective with photon flux greater than 1018/s, which results in 10 % reduction for 24 h irradiation. However, transmutation with photon flux 2 x 1012/s, which is achievable with present maximum accelerator systems, is not effective, and the reduction rate is approximately two orders of magnitude less than the natural decay rate.

Nuclear transmutation with laser Compton scattering can transmute selectively a medium mass nuclide AX into A-1X, and its reduction rate is independent of isotopes. Because the transmutation with laser Compton scattering can almost exclusively generate desired nuclides, this method will be useful for the generation of isotopes for medicine [1].

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.

Daisuke Ito, Kazuki Hase, and Yasushi Saito

Abstract In relationship to a solid target cooling system of an accelerator-driven system (ADS), wettability effect on boiling heat transfer has been experimentally investigated by irradiation with ultraviolet and gamma rays (y-rays). The experi­mental apparatus consists of a copper heater block, a rectangular container, and a thermostat bath. Two copper heater blocks were fabricated: one is for radiation — induced surface activation (RISA) and the other is for photoelectric reaction by ultraviolet whose heat transfer surface is coated by a TiO2 film. These copper heater blocks were irradiated by ultraviolet or by y-rays to change the surface wettability. Boiling heat transfer under subcooling conditions was measured before and after the irradiations to study the wettability effect. Experimental results show that nucleate boiling curves are shifted to the higher wall superheated side with the irradiated surface because of the decrease of the active nucleation sites. Heat transfer enhancement was found in both the critical heat flux and microbubble emission boiling (MEB) regions under these experimental conditions.

Keywords Microbubble emission boiling • Photocatalysis • Proton beam • Radiation-induced surface activation • Surface wettability

10.1 Introduction

An accelerator-driven system (ADS) is a hybrid-type nuclear system consisting of a proton accelerator, a spallation target, and a subcritical assembly in which high- energy particles and high heat density are generated in the target and subcritical assembly by the spallation and fission reactions. Lead-bismuth is considered the leading candidate for the liquid-metal spallation target for nuclear transmutation

D. Ito (*) • Y. Saito

Research Reactor Institute, Kyoto University, 2-1010 Asashiro-nishi, Kumatori-cho, Sennan-gun, Osaka 590-0494, Japan e-mail: itod@rri. kyoto-u. ac. jp

K. Hase

Power Systems Company, Toshiba Corporation, Kawasaki, 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_10

[1—3], whereas a solid target such as tungsten or tantalum should be also developed for a water-cooled ADS neutron source [4, 5].

High-energy radiation affects the surface wettability and boiling heat transfer of the solid target. Wettability on a solid surface can be changed by using ultraviolet radiation or y-rays, and recently the authors have found that the surface wettability can be also changed by proton-beam irradiation [6]. Applying the wettability change resulting from ultraviolet irradiation to titanium dioxide (TiO2), heat trans­fer research has been carried out to evaluate the wettability effect [7]. In addition, radiation-induced surface activation (RISA) enhances the surface wettability by irradiating a metal oxide layer with y-rays. Takamasa et al. [8] have applied RISA to heat-transfer experiments and reported that boiling heat transfer could be enhanced by changing the wettability of the heating surface. However, there has been no research to investigate surface wettability effect on boiling heat transfer at a solid target cooling system, where microbubble emission boiling (MEB) [9] might occur. MEB can take place when the heat transfer area is small (about 1 cm2) with subcooling conditions. In the target cooling system, the target should be cooled by subcooled water, and the heat-transfer area can be small when the proton beam is focused to a small area. Thus, MEB should be investigated for thermal hydraulic design and safety analysis of the solid target system, and also the effect of wettability on boiling heat transfer should be studied.

The purpose of this study is to investigate wettability change by ultraviolet, y-ray, and proton beam and to study the wettability effect on subcooled boiling heat transfer with a small heat-transfer area, and finally to obtain knowledge on the heat — transfer mechanism of the MEB phenomena.

The numerical validation for application of the self-indication method is discussed in this section. In the experiment, the transmitted neutron spectrum from the sample is measured via resonance reactions in the indicator whereas the reaction rates in the indicator are shown by numerical calculation. If the energy boundaries are made to have a finer division, the numerical result of the resonance reaction will have the same peak with dips as the measured data.

Figure 4.3 shows the 239Pu absorption rate yield by the present method (red) and the transmitted neutron spectrum by the conventional method (blue). The sample is the fresh (no burn-up) MOX pellet. Using the present method, one can easily obtain resonance absorption by 239Pu. On the other hand, the transmitted neutron spectrum has many dips caused by resonance reaction of the other nuclides. Thus, if the sample is a burn-up pellet, it is difficult to quantify and identify by using the conventional method.

A numerical result to identify 129I in the MOX pellet is described. The burn-up of the MOX pellet is 20 GWd/t. 129I has only four resonances in the energy region of

0. 1-100 eV: the resonance peaks are 41, 73, 75, and 97 eV. The transmitted neutrons are easily obtained via 129I resonance absorption reactions in the indicator by the present method (Fig. 4.4).

Using the self-indication method, one cannot prepare a pure indicator to identify and quantify a target nuclide in a sample. Therefore, it is necessary to validate the application of the present method using an impure indicator. Figure 4.5 shows the numerical result of 239Pu fission yield in the indicator, which has impure plutonium. The sample is a fresh MOX pellet, and the plutonium vector in the indicator is 239Pu = 98.57 w/o, 239Pu = 1.38 w/o, and 240Pu = 0.05 w/o. In Fig. 4.5, the red line is a pure 239Pu indicator, and the blue line shows that an indicator employed impure plutonium. Even in this case, as well as the result of using the pure 239Pu as the resonance absorption in indicator is observed, it is shown to quantify and identify 239Pu in the sample.

1.0E+00 1.0E+01 1.0E+02

Energy (eV)

Energy (eV)

Next, the applicability of the present method for fuel debris in Fukushima Daiichi NPP is examined. The fuel debris in Fukushima Daiichi NPP contains highly concentrated B-10, which has a large neutron absorption cross section. Thus, numerical validation of the present method and the conventional neutron transmission method for the sample with B-10 were carried out. The burn-up of the

sample is 30 GWd/t. Using the transmission neutron method, it is difficult to obtain the dips caused by resonance reaction (Fig. 4.6) because neutron absorption by B-10 has a large contribution in the sample. On the other hand, one can obtain the neutron absorption rate yield in an indicator by the present method although the signal of the neutron is decreased (Fig. 4.7).

4.2 Conclusion

Numerical validation for application of the self-indication method has been carried out. As a result, the self-indication method is shown to have a better S/N than the neutron transmission method to quantify the amount of target nuclides.

The present method can be applied to identify and quantify a nuclide that has a small resonance, i. e., 129I, and it is shown that one can measure an intended signal with good S/N by using an impure indicator. In addition, if the sample contains a highly concentrated neutron absorber, one can identify and quantify the target nuclide by using the self-indication method. Thus, the self-indication method can be applied to analyze the fuel debris in Fukushima Daiichi NPP.

Acknowledgments This work was supported by JSPS KAKENHI Grant Number 24760714.

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.