11.2.1 Four-Sensor Probe

The four-sensor probe [2] used in this study consists of a central front sensor and three peripheral rear sensors (Fig. 11.1a). Tungsten acupuncture needles with a maximum diameter of 0.1 mm were coated with epoxy resin varnish except the tip; the diameter of the tip is less than 1 pm. The insulated needles were inserted into a

a b

seven-bore insulating tube made of Al2O3. The output signals were acquired at a sampling frequency of 10 kHz and then processed on a PC.

Presentation of MYRRHA and Its Role in the European P&T Strategy

Hamid Ait Abderrahim

Abstract MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is an experimental accelerator-driven system (ADS) currently being developed at SCK*CEN for replacement of material testing reactor BR2. The MYRRHA facility is conceived as a flexible fast-spectrum irradiation facility that is able to run in both subcritical and critical modes. The applications catalogue of MYRRHA includes fuel developments for innovative reactor systems, material developments for GEN IV systems and fusion reactors, doped silicon production, radioisotope production, and fundamental science applications, thanks to the high — power proton accelerator. Next to these applications, MYRRHA will demonstrate the ADS full concept by coupling a high-power proton accelerator, a multi­megawatts spallation target, and a subcritical reactor at reasonable power level to allow operational feedback, scalable to an industrial demonstrator, and to allow the study of efficient transmutation of high-level nuclear waste. Because MYRRHA is based on the heavy liquid metal technology, namely lead-bismuth eutectic (LBE), it will be able to significantly contribute to the development of Lead Fast Reactor (LFR) technology and will have the role of European Technology Pilot Plant in the roadmap for LFR. The current design of the MYRRHA ADS and its ability to contribute to the European Commission strategy for high-level waste management through Partitioning and Transmutation (P&T) are discussed in this chapter.

Keywords ADS • HLW Management • MYRRHA • P&T

H. A. Abderrahim (*)

SCK*CEN, Boeretang 200, 2400 Mol, Belgium e-mail: haitabde@sckcen. be; myrrha@sckcen. be

© The Author(s) 2015

K. Nakajima (ed.), Nuclear Back-end and Transmutation Technology for Waste Disposal, DOI 10.1007/978-4-431-55111-9_7

1.1 Introduction

When concerned with energy, one cannot avoid considering geostrategic questions and the international political situation. Indeed, major armed conflicts in the world in past decades are taking place in major fossil energy production countries or on the major roads connecting places of great production with those of large consump­tion. Therefore, Europe is very concerned about the security of its supply in terms of energy, especially when considering the limited energy fossil resources in the European Union (EU). As such, nuclear power remains a major energy source in the EU.

Presently, the EU relies, for 30 % of its electric power production, on generation II-III fission nuclear reactors, leading to the annual production of 2,500 t/year of used fuel, containing 25 t plutonium, and high-level wastes (HLW) such as 3.5 t of minor actinides (MA), namely, neptunium (Np), americium (Am), and curium (Cm), and 3 t of long-lived fission products (LLFPs). These MA and LLFP stocks need to be managed in an appropriate way. The reprocessing of used fuel (closed fuel cycle) followed by geological disposal, or direct geological disposal (open fuel cycle), are today the envisaged solutions in Europe, depending on national fuel cycle options and waste management policies. The required time scale for geolog­ical disposal exceeds our accumulated technological knowledge, and this remains the main concern of the public. Partitioning and Transmutation (P&T) has been pointed out in numerous studies as the strategy that can relax constraints on geological disposal and reduce the monitoring period to technological and man­ageable time scales. Therefore, a special effort is ongoing in Europe and beyond to integrate P&T in advanced fuel cycles and advanced options for HLW manage­ment. Transmutation based on critical or subcritical fast-spectrum transmuters should be evaluated to assess the technical and economic feasibility of this waste management option, which could ease the development of a deep geological storage.

Despite diverse strategies and policies pursued by European Member States concerning nuclear power and the envisaged fuel cycle policy ranging from the once-through without reprocessing to the double-strata fuel cycle ending with ADS as the ultimate burner or generation IV (Gen-IV) fast critical reactors multi­recycling all transuranic (TRUs), P&T requires an integrated effort at the European and even worldwide level. Even when considering the phase-out of nuclear energy, the combination of P&T and a dedicated burner such as ADS technologies, at a European scale, would allow meeting the objectives of both types of countries, those phasing out nuclear energy as well as countries favoring the continuation of nuclear energy development toward the deployment of new fast — spectrum systems.

The concept of partitioning and transmutation has three main goals: reduction of the radiological hazard associated with spent fuel by reducing the inventory of minor actinides, reduction of the time interval required to reach the radiotoxicity of

natural uranium, and reduction of the heat load of the HLW packages to be stored in geological disposal, leading to its efficient use.

Transmutation of high-level radioactive elements with a long half-life present in the nuclear waste reduces the radiological impact of the actinides (such as ameri­cium, curium, and neptunium) and fission products. The time scale (Fig. 7.1) needed for the radiotoxicity of the waste to drop to the level of natural uranium will be reduced from a ‘geological’ value (500,000 to 1 million years) to a value that is comparable to that of human activities (several hundreds of years) [1­3]. During transmutation, the nuclei of the actinides are fissioned into shorter — lived fission products.

To transmute the minor actinides in an efficient way, high intensity and high energy neutron fluences are necessary. Therefore, only nuclear fast fission reactors, being critical or subcritical, can be utilized.

If the aim is to transmute large amounts of minor actinides in the dedicated transmuter then it is necessary to use an accelerator-driven system. The subcriticality is mandatory because of the smaller delayed neutron fraction within the minor actinides (0.01-0.1 %) compared to uranium-235 (0.7 %) to allow the criticality variation control.

After nearly 20 years of basic research funded by national programs and EURATOM framework programs, the research community needs to be able to quantify indicators for decision makers, such as the proportion of waste to be channeled to this mode of management, but also issues related to safety, radiation protection, transport, secondary wastes, costs, and scheduling.

From 2005, the research community on P&T within the EU started structuring its research toward a more integrated approach. This effort resulted, during the FP6, into two large integrated projects, namely, EUROPART dealing with partitioning, and EUROTRANS dealing with accelerator driven system (ADS), design for transmutation, development of advanced fuel for transmutation, R&D activities related to the heavy liquid metal technology, innovative structural materials, and nuclear data measurement. This approach resulted in a European strategy, the so-called four building blocks at engineering level for P&T, as given next. The implementation of P&T of a large part of the high-level nuclear wastes in Europe needs the demonstration of its feasibility at an “engineering” level. The respective R&D activities could be arranged in these four “building blocks,” as listed next:

1. Demonstration of the capability to process a sizable amount of spent fuel from commercial LWRs to separate plutonium (Pu), uranium (U), and minor actinides (MA),

2. Demonstration of the capability to fabricate, at a semi-industrial level, the dedicated fuel needed to load in a dedicated transmuter (JRC-ITU)

3. Design and construction of one or more dedicated transmuters

4. Provision of a specific installation for processing of the dedicated fuel unloaded from the transmuter, which can be of a different type than that used to process the original spent fuel unloaded from commercial power plants, together with the fabrication of new dedicated fuel

These “blocks will” result in identification of the costs and benefits of partitioning and transmutation for European society.

Laser Compton scattering is a method to obtain high-energy photons by laser photons backscattered off energetic GeV electrons. In the case of head-on collision

between relativistic electrons and laser photons, the energy of scattered photons is given by

where y = Ee/me is the Lorentz factor of the electron beam with energy Ee, me is the rest mass of the electron, El is the energy of the laser photon, and в is the scattering angle. From Eq. (1.4), the energy of the scattered photon is maximum at в = 0, and it depends on the energy of incident electrons and photons. The minimum energy of the scattered photon can be fixed by controlling в with collimators.

The scattering cross section of laser Compton scattering is given by the Klein— Nishina formula:

dNY do

Ny dEY dEY const.

Y Y dEY Y dEY

(1.5)

To have a situation in which (y, n) reactions occur, the photon beam with energy at B(n) < Ey < B(2n) is desired. In case of the free electron laser, we may assume/ expect to get the total photon flux NY « 2 x 1012/s/500mA for Ee = 1.2 GeV [1]. From Eq. (1.4), with Ee = 1.2 GeV and El = 0.7 eV, we obtain the maximum photon energy of Ey = 15 MeV at в = 0, which is equal to B(2n) for 137Cs. Figure 1.3 shows the calculated y-ray spectrum generated by laser Compton scattering using Eq. (1.5), where the total photon flux with energy from 0 to B(2n) is NY « 2 x 1012/s.

From Fig. 1.3, we can see that about half the total scattered photons are in B(n) < Ey < B(2n) and contribute to generate the (y, n) reactions for 137Cs. In contrast, for the Bremsstrahlung that is usually used to generate high-energy photons, the photon intensity decreases rapidly as the photon energy increases, and only a small part of the high-energy tail is available for (Y, n) reactions [11].

In the ADS with 100 MeV protons (Fig. 9.3), the fuel rod was composed of a thorium metal plate and a polyethylene (PE), graphite (Gr), or beryllium (Be) moderator arranged in the A-core. Other components were selected from HEU and natural uranium (NU; 2 x 2 x 1/8 in.) plates. The cores were composed of Th-PE (Fig. 9.4), Th-Gr, Th-Be, Th-HEU-PE, and NU-PE, according to a selection of moderator materials: PE, Gr, Be, HEU-PE, and NU-PE, respectively, and spallation neutrons were generated outside the core after injection onto the tungsten target. The thorium-loaded ADS experiments were conducted especially to investigate the relative influence of different neutron profiles on capture reactions of 232Th and 238U: the reaction of 238U was taken as reference data for evaluating the validity of 232Th capture cross sections.

1/8”Th (1/8”NU) Unit cell (3.18 mm) (15.88 mm) Al plate (20.00mm)

/_________

Fig. 9.4 Side view of Th-PE fuel assembly (TP) in thorium-loaded ADS core in Fig. 9.3

The main parameters of the proton beams were 100 MeV energy, 0.3 nA intensity, 20 Hz pulsed frequency, 100 ns pulsed width, and 40-mm-diameter spot size at the tungsten target (50 mm diameter and 9 mm thick). The level of the neutron yield generated at the target was around 1.0 x 107 1/s by the injection of 100 MeV protons onto the tungsten target.

Prompt and delayed neutron behavior was monitored by placing three 3He detectors (20 mm diameter and 300 mm long) at three locations. Throughout the time evolution of the prompt and delayed neutrons, the prompt neutron decay constant was deduced by least-squares fitting to an exponential function over the optimal duration. Subcriticality was deduced by the extrapolated area ratio method [13] on the basis of prompt and delayed neutron behaviors. For 100 MeV protons, neutron detectors (3He detectors: #1, #2, and #3) were set at three locations.

In this work, we proposed a new concept of the “self-indication method” as a complementary nondestructive assay for the fuel debris of Fukushima Daiichi NPP. We carried out experimental validation for application of the self-indication method. It was confirmed that the area density (thickness of the target nuclide) can be determined within 3 % accuracy by simple area analysis without a resonance fitting process. Moreover, it was experimentally shown that the contribution from the other nuclide can be remarkably suppressed by applying the self-indication method. The self-indication method combined with the TOF technique will be a useful tool for nondestructive assaying of the distribution of nuclear material in the melted fuel debris, which contains many impurities and has high activities.

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.

References

1. Postma H et al (2001) Radioanal Nucl Chem 248:115-120

2. Behrens JW, Johnson RG, Schrack RA (1984) Nucl Technol 67:162-168

3. Kiyanagi Y et al (2005) Measurement of eV-region pulse shapes of neutrons from KENS thermal neutron source by a neutron resonance absorption method. J Nucl Sci Technol 42 (3):263-266

4. Pietropaolo A et al (2010) A neutron resonance capture analysis experimental station at the ISIS spallation source. Appl Spectrosc 64(9):1068-1071

5. Kobayashi K et al (1987) KURRI-Linac as a neutron source for irradiation. Annu Rep Res Reactor Inst Kyoto Univ 22:142

6. Yamamoto S et al (1996) Application of BGO scintillators to absolute measurement of neutron capture cross sections between 0.01eV to 10eV. J Nucl Sci Technol 33:815

7. Shibata K et al (2011) JENDL-4.0: a new library for nuclear science and engineering. J Nucl Sci Technol 48:1

The schematic of the electromagnetic probe [3, 4] used in this study is shown in Fig. 11.1b. The probe consists of a SmCo magnet, electrode wires, and a stainless steel jacket. Because a small cylindrical magnet with a diameter of 2 mm was used to miniaturize probe size, the induced potential between the electrodes at the tip of the probe was rather small. Therefore, the detected signal was amplified by a low-noise pre-amplifier and a DC amplifier. The signals digitized by an A/D converter were processed on a PC. The sampling frequency was 10 kHz.

The principle of the electromagnetic probe is based on Faraday’s law. When the conducting fluid passes across a magnetic field, potential is induced in a direction normal to the magnetic field and the fluid velocity. Here, the induced potential is proportional to the velocity. In this study, the calibration of the electromagnetic probe was carried out using a rigid rotating setup (Fig. 11.2a) that consists of a cylindrical tank, a rotating system, and a heater. The tank was filled with LBE and rotated at a constant speed. The probe was inserted into the molten LBE rotating rigidly in the tank. The voltage corresponding to the tangential velocity component was measured, and this calibration was performed for all probes used in this study. Typical calibration results are shown in Fig. 11.2b.

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK*CEN. MYRRHA is able to work both in subcritical (ADS) and in critical mode. In this way, MYRRHA targets the following applications catalogue:

• To demonstrate the ADS full concept by coupling the three components (accel­erator, spallation target, and subcritical reactor) at reasonable power level (50­100 MWth) to allow operation feedback, scalable to an industrial demonstrator;

• To allow the study of the efficient technological transmutation of high-level nuclear waste, in particular, minor actinides that would require high fast flux intensity ^>0.75MeV = 1015 n/cm2 s);

• To be operated as a flexible fast-spectrum irradiation facility allowing for

— Fuel developments for innovative reactor systems, which need irradiation rigs with a representative flux spectrum, a representative irradiation temperature, and high total flux levels (0tot = 5 • 1014 to 1015 n/cm2 s); the main target will be fast-spectrum GEN IV systems, which require fast-spectrum conditions;

— Material developments for GEN IV systems, which need large irradiation volumes with high uniform fast flux level (Ф>1 MeV = 1~5-1014 n/cm2 s) in various irradiation positions, representative irradiation temperature, and rep­resentative neutron spectrum conditions; the main target will be fast- spectrum GEN IV systems;

— Material developments for fusion reactors, which need also large irradiation volumes with high constant fast flux level (Ф>і MeV = 1 ~ 5 • 1014 n/cm2 s), a representative irradiation temperature, and a representative ratio appm He/dpa(Fe) = 10;

— Radioisotope production for medical and industrial applications by

• Holding a backup role for classical medical radioisotopes;

• Focusing on R&D and production of radioisotopes requiring very high thermal flux levels (Ф^ит^ = 2 to 3-1015 n/cm2 s) because of double­capture reactions;

— Industrial applications, such as Si-doping, need a thermal flux level depending on the desired irradiation time: for a flux level Ф^ит^ = 1013 n/ cm2 s, an irradiation time in the order of days is needed, and for a flux level of

thermai = 1014 n/cm2 s, an irradiation time in the order of hours is needed to obtain the required specifications.

Further in this section, we discuss some basic characteristics of the accelerator and of the core and primary system design.

When a target nucleus X is irradiated with a photon beam with energy Ey, it forms a compound nucleus, which releases one neutron and becomes its isotope X0. The reaction rate of X(Y, n)X0 at time t is given by

B(2n)

rX—X'(t) = J d£y dEx ffX! X, (Ey)nx(t)a, (1.6)

B(n) Y

where ffX —x>(Ey) is the reaction cross section of X(y, n)X’, nx( t) is the number of target nucleus per unit area at time t, and a is the attenuation factor of incident photons through a thick target. dNy/dEy is expressed with Eq. (1.5) and aX — X’ (Ey) is calculated from Eq. (1.2) using the TALYS code.

Figure 1.4 shows a calculation in which the photon beam is generated by the laser Compton scattering of 1.2 GeV electrons and 0.7 eV laser beams. We assume that the cylindrical target of 137Cs of 1 g is irradiated with a photon beam with energy B(n) < EY < B(2n) within a radius r « 0.8 mm at 2 m from the interaction point. When a target of 137Cs is irradiated with photons and is excited to GDR, the (Y, Y), (Y, n) and (y, 2n) reactions mainly occur. We consider 137Cs, 136Cs, 135Cs, and 134Cs as the isotopes generated by the transmutation. The numbers of these isotopes are expressed as

П137 (t + At) = П137 (t)e 7,137At — r137!136(t) — r137!135(t), (1.7)

П136(t + At) = n136(t)e 7136At + Г137 —^136(t) — Г136!135(t) — П36—134(t), (1.8)

n135 (t + At) = n 135 (t)e 7135At + r 137—135 (t) + r 136—135 (t) — r 135—134 C’1) . (1.9)

n134(t + At) = n134(t)e ^1344t + r 136—134(t) + r 135—134(t). (1.10)

One can calculate the number of each isotopes by solving these equations with the Runge-Kutta method.

To obtain information on the detector position dependence of the prompt neutron decay measurement, the neutron detectors were set at three positions as shown in Fig. 9.1: near the tungsten target [position of (17, D): 1/2-in.-diameter BF3 detec­tor]; and around the core [positions of (18, M) and (17, R): 1-in.-diameter 3He detectors]. The prompt and delayed neutron behaviors (Fig. 9.6) were

c

1000

о

£

c

з

100

c

о

з

CD

z

10

experimentally confirmed by observing the time evolution of neutron density in ADS, an exponential decay behavior and a slowly decreasing one, respectively. These behaviors clearly indicated that the neutron multiplication was caused by an external source: the sustainable nuclear chain reactions were induced in the sub­critical core by the spallation neutrons through the interaction of the tungsten target and the proton beams from the FFAG accelerator. In these kinetic experiments, the subcriticality was deduced from the prompt neutron decay constant by the extrap­olated area ratio method. The difference of measured results of 0.74 %Ak/k and 0.61 %Ak/k at the positions of (17, R) and (18, M) in Fig. 9.1, respectively, from the experimental evaluation of 0.77 %oAk/k, which was deduced from the combination of both the control rod worth by the rod drop method and its calibration curve by the positive period method, was within about 20 %. Note that the subcritical state was attained by a full insertion of C1, C2, and C3 control rods into the core.

Tadafumi Sano, Jun-ichi Hori, Yoshiyuki Takahashi, Hironobu Unesaki, and Ken Nakajima

Abstract To perform decommissioning of the Fukushima Daiichi NPP safely, it is very important to measure the components of the fuel debris. Therefore, a new nondestructive assay to identify and quantify the target nuclide in fuel debris using a pulsed-neutron source is under development in Kyoto University Research Reac­tor Institute.

We use the self-indication method for the nondestructive assay. This method is a neutron transmission method. The neutron transmission method is focused on resonance reactions (i. e., capture, fission) at the target nuclide. In the self-indication method, the transmitted neutrons from the sample are injected into an indicator. The indicator consists of a high-purity target nuclide. The transmitted neutrons are obtained by the time-of-flight (TOF) technique via resonance reactions in the indicator. The self-indication method has a high signal-to-noise (S/N) ratio com­pared to the conventional method.

In this study, numerical validation for the self-indication method to identify and quantify nuclides in a BWR-MOX pellet is described. The burn-up of the MOX pellet is 0 GWd/t, 10 GWd/t, 20 GWd/t, 30 GWd/t, 40 GWd/t, and 50 GWd/t. The 12-m measurement line in KUR-LINAC is simulated as a calculational geometry. Numerical calculations are carried out by continuous-energy Monte-Carlo code MVP2 with JENDL-4.0 as the nuclear data library. The burn-up calculations of the BWR-MOX pellet are performed by the deterministic neutronics code SARC 2006 with JENDL-4.0.

Numerical validation for application of the self-indication method is carried out. From the results, it is noted that the self-indication method has a good S/N ratio compared to the neutron transmission method for quantifying the amount of target nuclides in the fuel debris.

T. Sano (*) • J. Hori • Y. Takahashi • H. Unesaki • K. Nakajima Kyoto University Research Reactor Institute, 1010, Asashiro-nishi-2, Kumatori-cho, Sennan-gun, Osaka, Japan e-mail: t-sano@rri. kyoto-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_4

Keywords Burn-up • KUR-LINAC • MOX pellet • Nondestructive assay • Numer­ical validation • Resonance • Self-indication method

4.1 Introduction

To perform decommissioning of the Fukushima Daiichi NPP safely, it is very important to measure the components of the fuel debris. Therefore, a new nonde­structive assay to identify and quantify a target nuclide in the fuel debris using a pulsed-neutron source is under development in Kyoto University Research Reactor Institute.

We use the self-indication method for the nondestructive assay. This method is a neutron transmission method. The neutron transmission method is focused on resonance reactions (i. e., capture, fission) at the target nuclide. In the conventional neutron transmission method, a sample is irradiated by a pulsed-neutron beam and the energy distribution of transmitted neutrons from the sample is measured by the time-of-flight technique. Then, the target nuclide in the sample is identified and quantified by using the transmitted neutrons in the resonance energy region. This is a remarkably effective method to identify and quantify the target nuclide. However, if the energy spectrum of the transmitted neutron has many dips caused by reso­nance reactions of other nuclides, it is difficult to identify and quantify the target nuclide in the sample.

In the self-indication method, the transmitted neutrons from the sample are injected into an indicator, which consists of a high-purity target nuclide. The transmitted neutrons are obtained via resonance reactions in the indicator. The self-indication method has a high signal-to-noise (S/N) ratio compared to the conventional method.

In this chapter, numerical validation for application of the self-indication method is carried out. A calculational model and conditions are shown in Sect. 4.2 and the numerical results are shown in Sect. 4.3. From these results, some conclusions are drawn in Sect. 4.4.