Category Archives: Nuclear Back-end and Transmutation Technology for Waste Disposal

Measurement Activities by the Activation Method

Neutron capture cross sections were determined on the basis of Westcott’s conven­tion [1] by an activation method. The results for LLFPs [223] are listed in Table 5.1, and for MAs [2431] in Table 5.2, together with previously reported data. Here, the symbols oeff, o0, and /0 denote the effective cross section, the thermal neutron capture cross section, and the resonance integral, respectively; o0 is the cross section at the neutron energy of 25.3 meV.

Nuclear waste sometimes contains a large amount of stable nuclei having the same atomic number as that of long-lived fission products. These stable nuclei absorb thermal neutrons during the neutron irradiation of the nuclear waste and affect the neutron economics; the reaction rate of the target nuclei is reduced. Moreover, some of these stable nuclei breed more radioactive nuclei by the neutron capture process. It is also necessary for transmutation study to accurately estimate these influences caused by stable nuclei involved in the FP targets. The cross sections of the stable nuclei, such as I [14] and Cs [20], were also measured; the results are shown in Table 5.1.

Table 5.1 Results of thermal neutron capture cross sections and resonance integrals for long — lived fission products (LLFPs)

Nuclide (half-life)

Reported data (author, year)

JAEA data

137Cs (30 years)

°eff = 0.11 ± 0.03 b (Stupegia 1960 [2])

o0 = 0.25 ± 0.02 b

I0 = 0.36 ± 0.07 b (1990,1993,2000 [35])

90Sr (29 years)

oeff = 0.8 ± 0.5 b (Zeisel 1966 [6])

o0 = 10.1 ± 1.3/4.2 mb

I0<0.16b(1994 [7])

o0 = 10.1 ± 1.3 mb

I0 = 104 ± 16 mb (2001 [8])


(2.1 x 105 years)

o0 = 20 ± 2b

o0 = 22.9 ± 1.3 b

I0 = 186 ± 16 b (Lucas 1977 [9])

I0 = 398 ± 38 b (1995 [10])


(1.6 x 107 years)

o0 = 27 ± 2b

o0 = 30.3 ± 1.2 b

I0 = 36 ± 4 b (Eastwood 1958 [11])

I0 = 33.8 ± 1.4 b (1996 [12])

127I (Stable)

o0 = 4.7 ± 0.2 b

o0 = 6.40 ± 0.29 b

I0 = 109 ± 5 b (Friedman 1983 [13])

I0 = 162 ± 8 b (1999 [14])


(3 x 106 years)

o0 = 8.7 ± 0.5 b

o0 = 8.3 ± 0.3 b

I0 = 61.7 ± 2.3 b (Baerg 1958 [15])

I0 = 38.1 ± 2.6 b (1997 [16])

134Cs (2 years)

oeff = 134 ± 12 b (Bayly 1958 [17])

oeff = 141 ± 9 b (1999 [18])

133Cs (Stable)

o0 = 30.4 ± 0.8 b

o0 = 29.0 ± 1.0 b

I0 = 461 ± 25 b (Baerg 1960 [19])

I0 = 298 ± 16 b (1999 [20])


(1.2 x 103 years)

o0 = 9,140 ± 650 b

oeff = 3 ± 1 kb (2000 [22])

I0 = 1,140 ± 90 b (Masyanov 1993 [21])

o0 = 3.11 ± 0.82 kb

I0 = 10.0 ± 2.7 kb (2002 [23])

Table 5.2 Results of thermal neutron capture cross sections and resonance integrals for minor actinides (MAs)

Nuclide (half-life)

Reported data (author, year)

JAEA data


(2.14 x 106 years)

o0 = 158 ± 3 b

o0 = 141.7 ± 5.4 b

I0 = 652 ± 24 b (Kobayashi 1994 [24])

I0 = 862 ± 51 b (2003 [25])

o0 = 169 ± 6 b (2006 [26])

238Np (2.1 days)

No data

oeff = 479 ± 24 b (2004 [27])

241Am (432 years)

o0, g = 768 ± 58 b

°0, g = 628 ± 22 b

I0 g = 1,694 ± 146 b (Shinohara 1997 [28])

I0 g = 3.5 ± 0.3 kb (2007 [29])

243Am (7,370 years)

°0, m = 80 b

oeff = 174.0 ± 5.3 b (2006 [31])

o0, g = 84.3 b (Ice 1966 [30])

As seen in Table 5.1, the thermal cross section for 137Cs is about twice as large as the previous data reported by Stupegia [2]. As for 90Sr, its thermal cross section is

found to be much smaller than the data reported by Zeisel [6]. As seen in Table 5.2, the cross section of 238Np is obtained for the first time. Thus, the joint research activities of the Japan Atomic Energy Agency (JAEA) and universities have measured the cross sections for important LLFPs and MAs by the activation method.

Theory of Power Spectral Density in ADS

Another subcriticality measurement technique is the power spectral density method. This chapter focuses on the cross-power spectral density (CPSD), which is the Fourier transformation of a cross-correlation function between two neutron detector signals. In an infinite and homogeneous subcritical system where the energy and spatial dependence of the neutron is neglected, the CPSD is simply expressed as a function of frequency:

CPSD(m) / 1/(m2 + «2), (12.9)


where m = angular frequency. The CPSD in an ADS where the energy and spatial dependence is considered, however, is much more involved as

where i = л/—Г, mm = 2nm/T, and the subscripts C, UN, and CS have the same meanings as in the previous section. For the two systems in the previous section (large subcritical system and nearly critical system), Monte Carlo simulations were performed to obtain CPSDs between detectors 1 and 2 in Fig. 12.1. In the simula­tions, the pulse period is T = 0.05 s (20 Hz). The simulation result for the nearly critical system is compared with the theoretical one in Fig. 12.6. The results of the large subcritical system are shown by Yamamoto [15]. The theoretical results agree well with the Monte Carlo simulations. The uncorrelated component, CPSDUN(m), emerges only at the integer multiples of the pulse frequency as the Delta-function­like peaks. Thus, either of the correlated and uncorrelated components can be easily discriminated from the CPSD. Using Eq. (12.12), the uncorrelated and correlated components of the CPSD in the nearly critical system is decomposed into the mode components, shown in Figs. 12.7 and 12.8, respectively. In the correlated

3.5E+06 3.0E+06 g 2.5E+06 1 2.0E+06 1.5E+06 1.0E+06 5.0E+05

0. 0E+00 -5.0E+05

Fig. 12.7 Mode components of the uncorrelated component in the CPSD in the nearly critical system (real part)

component, the higher-order modes are negligibly small, and almost the whole of the CPSD is made up of the fundamental mode. The same condition holds for the large subcritical system. The higher-order mode effect in the correlated component is minor even in the large subcritical system. In the uncorrelated component, the higher-order mode effect is significant even in the nearly critical system. In the large subcritical system, the higher-order mode effect is much more significant. Thus, fitting the uncorrelated component to Eq. (12.9) yields an inaccurate a value unless the system is nearly critical. For example, in the large subcritical system we obtain a = 789 (s-1) for the true fundamental mode a value of 940 (s-1) [15] from the uncorrelated component. On the other hand, we obtain a = 900 (s-1) from the

Frequency (Hz)

Подпись: Fig. 12.8 Mode components of the correlated component in the CPSD in the nearly critical system (real part)

correlated component. In the nearly critical system, we obtain a = 172 (s-1) and 178 (s-1) from the uncorrelated and correlated component, respectively, for the true fundamental mode a value of 179 (s-1) [15].

12.3 Conclusions

In a subcriticality measurement for an ADS, the Feynman Y function in general appears as the sum of the correlated and uncorrelated components. The higher­mode effect in the correlated component is less significant than in the uncorrelated component. Thus, a relatively good approximation of the true fundamental mode a can be obtained by using the correlated component. However, it is not necessarily easy to separate the correlated component from the measured Feynman Y function. Considering the difficulty of separating the correlated component, the Feynman-a method is not always suitable as a subcriticality measurement technique for ADSs. In an ADS that is nearly critical, the uncorrelated component is very minor. Thus, by fitting the measured Feynman Y function to the correlated component, the fundamental mode a can be accurately estimated.

In a subcriticality measurement using the power spectral density method, the uncorrelated component emerges at the integer multiples of the pulse frequency as delta-function-like peaks. Thus, the uncorrelated component can be easily discrim­inated from the correlated component. The correlated component is less contami­nated by the higher-order modes. An approximate fundamental mode a can be obtained by fitting the Feynman Y function to the correlated component of the power spectral density. The use of the uncorrelated component is not always recommended, because the higher-order modes are more significant in the uncorrelated component.

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.

Nuclear Back-end and Transmutation Technology for Waste Disposal

On March 11, 2011, a massive earthquake and the resultant tsunami struck the Tohoku area of Japan, causing serious damage to the Fukushima Daiichi Nuclear Power Plant (NPP) and the release of a significant quantity of radionuclides into the surrounding environment. This accident underlined the necessity of establishing more comprehensive scientific research for promoting safety in nuclear technology. In this situation, the Kyoto University Research Reactor Institute (KURRI) established a new research program called the “KUR Research Program for Scien­tific Basis of Nuclear Safety” in 2012.

Nuclear safety study includes not only the prevention of nuclear accidents but also the safety measures after the accident from a wider point of view ensuring the safety of residents. A long time is needed for the improvement of the situation, but the social needs for the reinforcement of nuclear safety increases rapidly. The advancement of disaster prevention technology for natural disasters such as earthquakes and tsunamis, the reinforcement of measures to counter the effects of accidents, and the reinforcement of the safety management of spent fuels and radioactive wastes are demanded, not to mention the reinforcement of nuclear reactor safety. Also required are the underlying mechanism investigation and accurate assessment for the effect of radiation on the human body and life. As with all such premises, detailed inspection and analysis of the accident are indispensable.

In the Research Program for the Scientific Basis of Nuclear Safety, an annual series of international symposia was planned along with specific research activities. The first in the series of symposia, entitled “The International Symposium on Environmental Monitoring and Dose Estimation of Residents after Accident of TEPCO’s Fukushima Daiichi Nuclear Power Stations”, was held on December 14, 2012, concerning the radiological effects of the accident on the public, and covering a wide range of monitoring and dose assessment activities after the accident. Although the proceedings of the symposium had been published, a more comprehensive and conclusive book was published with open access at the requests of many people including residents near the accident site.

Following the first one, the second annual symposium in this series was held on November 28, 2013, dealing with nuclear back-end issues and the role of nuclear transmutation technology after the accident at TEPCO’s Fukushima Daiichi NPP. The accident has called upon us to focus our attention on the large amount of spent nuclear fuels stored in NPPs as well as on the impacts of the accident. In fact, public anxiety regarding the treatment and disposal of high-level radioactive wastes which require long-term control is now growing, while the government policy on the back-end of the nuclear fuel cycle is unpredictable in the aftermath of the accident. The issues are not simply technical, they are critically important not only for dealing with the accident but also for pursuing nuclear energy production in the world.

This publication summarizes the current status of the back-end issues and of research and development on nuclear transmutation technology for radioactive waste management. It is expected to contribute to better understanding and further discussion of the issues.

On behalf of KURRI, I wish to thank all the contributors to this book as well as the reviewers. KURRI hopes that this publication will promote further progress in nuclear safety research and will contribute to the reduction of public anxiety after the accident.

Kyoto University Research Reactor Institute Hirotake Moriyama

Kyoto, Japan

Outline of TEF-P

Several neutronic experiments for ADS have been performed in both Europe [6, 7] and Japan. In Japan, subcritical experiments with fast neutron spectrum core were performed at the Fast Critical Assembly (FCA) in JAEA/Tokai, and subcritical experiments with thermal subcritical core driven by 100 MeV protons are performed at Kyoto University Research Reactor Institute. Many experimental studies also have been performed on the neutronics of the spallation neutron source with various target materials such as lead, tungsten, mercury, and uranium. These experiments for spallation targets are also useful to validate the neutronic charac­teristics of ADS. However, there are no experiments combined with a spallation source installed inside the subcritical fast-neutron core. The purpose of the TEF-P is divided roughly into three subjects: (1) reactor physics aspects of the subcritical core driven by a spallation source, (2) demonstration of the controllability of the subcritical core including power control by the proton beam power adjustment, and

(3) investigation of the transmutation performance of the subcritical core using a certain amount of MA and LLFP.

TEF-P is designed with referring to FCA, the horizontal table-split type critical assembly with a rectangular lattice matrix. In this concept, the plate-type fuel for FCA with various simulation materials such as lead and sodium for coolant, tungsten for solid target, ZrH for moderator, B4C for absorber, and AlN for simulating nitride fuel can be commonly used at TEF-P. Therefore, previous experiments can be correlated with TEF-P experiments. The proton beam will be introduced horizontally at the center of the fixed half assembly, and various kinds of spallation targets can be installed at various axial position of the radial center of the subcritical core. Application of MA fuel is one of the promising characteristics of TEF-P. Installation of a partial mock-up region of MA fuel with air cooling is considered to measure the physics parameters of the transmutation system. R&D to utilize MA fuel by remote handling systems is under way.

Development of a y-Ray Spectrometer for NRCA/PGA

The y-ray background from debris is expected to be strong. The strongest radioac­tive isotope in a MF of the TMI-2 accident was 137Cs, which ranged from 106 to 3 x 108 Bq/g [14]. The energies of the prominent у rays from nuclei listed in Table 2.1 is larger than the 661 keV y-rays from 137Cs, except for 10B. Accordingly, most of the measurements of the NRCA/PGA will not have interference with the y-rays from 137Cs. On the other hand, the Compton edge of the 661 keV y-rays surely overlaps with the 478-keV y-ray peak originating the 10B(n, ay)7Li reaction.

Подпись: Al cases & PMTs


Подпись: Y ray source position
Подпись: T#
Подпись: / ^ LaBr3 crystals

The y-ray spectrometer used for NRCA/PGA, therefore, requires properties of not only high-energy resolution and fast timing response, but also a high peak-to — Compton ratio.

To satisfy these requirements, a well-type spectrometer made of LaBr3 crystal has been proposed [2, 15]. In a study based on Monte Carlo simulations of a well — type LaBr3 spectrometer [15, 16], the Compton edge was successfully reduced by an order of magnitude. Such reduction enables us to roughly quantify 10B in a sample, even in the presence of high background y-rays from 137Cs.

A prototype LaBr3 y-ray spectrometer has been designed (Fig. 2.2). Because of the technical difficulty of producing a crystal with a well, the spectrometer is made of several detectors: a cylindrical detector and four square pillar detectors. The cylindrical crystal is 120 mm in diameter and 127 mm in length; each square pillar crystal is 50.8 x 50.8 x 76.2 mm. An arrangement of the detector pillars opens a square channel of 20 x 20 mm for the passage of collimated y-rays from the samples. This spectrometer will be tested soon.

Effect of Irradiations on Surface Wettability

Figure 10.4a-c shows the wettability change from ultraviolet, y-ray, and proton-beam irradiation, respectively. The horizontal axis denotes an integrated irradiation dose or irradiation time and the vertical axis denotes the measured contact angle. As shown in these figures, the contact angle decreases with the irradiation dose. In these experi­ments, the ambient effect is also studied during the irradiations, which are performed in air or water. As shown in Fig. 10.4a, the ambient effect on the contact angle change is not obvious in the ultraviolet irradiation. However, the ambient effect is very distinct both in the y-ray and the proton-beam irradiations. It is suggested that the wettability enhancement by the radiations may be attributed to the radiolysis of water.

Measurement Activities at J-PARC/MLF/ANNRI

A new experimental apparatus called the accurate neutron nucleus reaction mea­surement instrument (ANNRI) has been constructed on the beam line no. 4 (BL04) of the MLF in the J-PARC. The ANNRI has two detector systems. One of them is a large Ge detector array, which consists of two cluster-Ge detectors, eight coaxial­shaped Ge detectors, and BGO Compton suppression detectors; the other is a large NaI(Tl) spectrometer (Fig. 5.3). The ANNRI has an advantage for neutron cross­section measurements because the MLF facility can provide the strongest neutron intensity in the world.

The neutron capture cross sections of 237Np [32, 33], 241Am [34], 244Cm [35], 93Zr [36], 99Tc [37], and 107Pd [38] have been measured relative to the 10B(n, ay) standard cross section by the TOF method. Some highlights of results obtained in our research activities are shown in Fig. 5.4 for 237Np and in Fig. 5.5 for 93Zr.


Fig. 5.3 A new experimental apparatus called the accurate neutron nucleus reaction measurement instrument (ANNRI). The cross-sectional view of ANNRI is shown in the upper panel, the spectrometer is on the left side, and the NaI(Tl) spectrometer is on the right side

image30This work

О Weston(1981) □ Esch(2008)

Neutron energy (ev)

Fig. 5.4 241Am cross section in neutron energy from 0.01 to 10 eV

The results obtained at the ANNRI are good agreement with the data reported by Weston (Fig. 5.4). The 93Zr cross sections in Fig. 5.5 present results greatly different from the evaluated data in the thermal neutron energy region. One finds that the present results support the value of the thermal cross section reported in 2007 [39].

5.2 Summary

This chapter described the JAEA research activities for the measurement of neutron capture cross sections for LLFPs and MAs by activation and neutron time-of-flight (TOF) methods. We summarized our results of the thermal neutron capture cross

image31Neutron energy (eV)


Подпись: Present JENDL-4.0 ENDF-B/V1I Macklin (1985) Nakamura (2007) Pomerance (1955)
Подпись: c о '+■> о <D CO to to О v. О a; L. 3 -H a to о c о

93Zr cross section (tentative data) together with the evaluated data

section and the resonance integral for some of the important LLFPs and MAs by the activation method.

Operation of a new experimental apparatus called the accurate neutron nucleus reaction measurement instrument (ANNRI) in the MLF at J-PARC has been started for neutron capture cross-section measurements of MAs and LLFPs. Some of the highlights of our results have been shown here.

Acknowledgments The authors thank the staff at Kyoto University Reactor Institute, Rikkyo University Reactor and JRR-3M. A part of this work has been carried out under the Visiting Researcher’s Program of the Research Reactor Institute, Kyoto University. Moreover, the authors appreciate the accelerator staff of J-PARC for their operation of the accelerator.

This work is supported by JSPS KAKENHI (22226016).

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.


The accident at the Fukushima Daiichi Nuclear Power Plant (NPP), which occurred on March 11, 2011, has caused us to focus our attention on a large amount of spent nuclear fuels stored in NPPs. In addition, public anxiety regarding the treatment and disposal of high-level radioactive wastes that require long-term control is growing. The Japanese policy on the back-end of the nuclear fuel cycle is still unpredictable in the aftermath of the accident; moreover, these back-end issues are inevitable as long as nuclear energy is used. Therefore, research and development for enhancing the safety of various processes involved in nuclear energy production is being actively pursued worldwide. In particular, the nuclear transmutation technology— employed for reducing the toxicity of highly radioactive wastes—has been drawing significant attention.

In the KUR Research Program for Scientific Basis of Nuclear Safety, the Kyoto University Research Reactor Institute organized an international symposium “Nuclear back-end issues and the role of nuclear transmutation technology after the accident of TEPCO’s Fukushima Daiichi Nuclear Power Stations” in November 28th, 2013. Under such circumstances in which nuclear back-end issues and the role of nuclear transmutation technology after the accident at TEPCO’s Fukushima Daiichi NPP is gaining greater concern, this timely publication highlights the following topics: (1) Development of accelerator-driven systems (ADS), which is a brand-new reactor concept for transmutation of highly radioactive wastes; (2) Nuclear reactor systems from the point of view of the nuclear fuel cycle. How to reduce nuclear wastes or how to treat them including the debris from TEPCO’s Fukushima nuclear power stations is discussed; and (3) Environmental radioactiv­ity, radioactive waste treatment, and geological disposal policy.

State-of-the-art technologies for overall back-end issues of the nuclear fuel cycle as well as the technologies of transmutation are presented here. The chapter authors are actively involved in the development of ADSs and transmutation-related


Подпись: Prefacetechnologies. The future of the back-end issues is very uncertain after the accident in Fukushima Daiichi NPP, and this book provides an opportunity for readers to consider the future direction of those issues.

Sennan-gun, Osaka, Japan Ken Nakajima

Design of Spallation Target for TEF-T

To evaluate the feasibility of a designed beam window of TEF target, numerical analysis with a three-dimensional (3D) model was performed. The analysis was done by considering the current density and shape of the incident proton beam to the target and the thermal fluid behavior of Pb-Bi around the beam window as a function of flow rate and inlet temperature. The thickness of the beam window is also considered from 2 to 3 mm. After the temperature distribution analysis, structural strength of the beam window is determined to evaluate soundness of the target. A concave shape beam window was used for this analysis. The prototype design of the beam window for TEF target system is shown in Fig. 8.2.

The material of the beam window would be a type 316 stainless steel. The concave section in the center part of the target was connected to the convex section in the terminal part, and then it was connected to the straight tube. A straight tube part has coaxially arranged annular and tube-type channels. The inner diameters of the outside tube and inside tube were set to 150 and 105 mm, respectively. The total length of the analysis region was 600 mm, which corresponds to an effective target depth for the 400 MeV proton. An irradiation sample holder, which was installed in the inner tube, holds eight irradiation specimens in the horizontal direction. The size of each specimen was 40 x 145 x 2 mm. The rectification lattice having the aperture of the plural squares type was installed at the front end of the sample holder. A slit 2 mm in width was arranged along the side of the rectification lattice to cool the sample holder by flowing Pb-Bi.

The thermal-fluid behavior of the target was analyzed by the STAR-CD. The quarter-part model was set to tetra metric type and the divided face was set to a reflected image condition. At first, Pb-Bi flowed through the annular region and joined in the center of the beam window, and then, turned over and flowed in the inner tube after having passed a rectification lattice and an irradiation sample. In a default condition, flow rate at the inlet of annulus region was set to 1 l/s, and this was equivalent to the flow velocity of 0.125 m/s. Because the Pb-Bi flow forms a

Fig. 8.2 Prototype LBE spallation target for TEF-T

image43complicated turbulent flow, the standard k-є model for high Re number type was used for a turbulence model. Heat deposition distribution by the primary proton beam, which was calculated by a hadronic cascade code PHITS [8], was used for the analysis. The internal pressure to the inside of the beam window was set to

0. 3 MPa in consideration of the flowing Pb-Bi and the cover gas. On the outer side of the beam window and the border of the atmosphere, release of the radiant heat was considered. Embrittlement of the structural materials by irradiation was not considered.

The analyses were performed by changing flow rates from 1 to 4 l/s. In each case, a dead region was commonly formed in the center of the inside of the beam window. The maximum velocity of Pb-Bi was confirmed at the rectification lattice part and was approximately 1.2 m/s in the case of the inlet flow rate of 1 l/s. When the inlet flow rate increased to 4 l/s, the maximum velocity in the target reached

4.8 m/s, which is too high to apply to the Pb-Bi target. The maximum temperature is 544 °C in the case of a 3-mm-thick window. The peak temperature can be decreased to 477 °C in the case of 2-mm-thick window. The temperature differences between outside and inside at the center of the window were 65 and 37 °C in the case of the 3-mm-thick window and the 2-mm-thick window, respectively. From these results, it was determined that a condition of 2 mm was desirable.

Based on the results provided by STAR-CD, analysis to verify the feasibility of the beam window was performed by ABAQUS code. The operating conditions for the first stage of material irradiation in TEF were decided by a result of the analysis on each condition. The temperature and thermal stress for the steady state were estimated using ABAQUS code, the computational code for the finite-element method. In the ABAQUS code, only a beam window was modeled as the cylinder-slab geometry. From the analysis result for the 2-mm-thick window, the stress strength reached the maximum value of 190 MPa on the outer surface of the beam window. When the maximum temperature of the beam window is adopted to 470 °C from the result of STAR-CD, maximum stress is lower than the tolerance level of the materials for fast reactor, and hence, the feasibility of a designed beam window was confirmed.

8.4 Conclusion

To perform the design study for the transmutation system of long-lived nuclides, the construction of TEF, which consists of two buildings, TEF-T and TEF-P, is proposed under the J-PARC Project. According to the current construction sched­ule, TEF-T will be built at the first phase and TEF-P will be constructed at the latter phase. Licensing procedures for TEF-P construction will be processed simulta­neously with TEF-T construction.

TEF-T is a facility to prepare the database for engineering design of an ADS using a 400 MeV-250 kW proton beam and the Pb-Bi spallation target. The purposes of TEF-T are R&D for the structural strength of the beam window, which is irradiated by both high-energy protons and neutrons, compatibility of the structural material with flowing liquid Pb-Bi, and operation of the high-power spallation target. Several kinds of target head can be installed according to the experimental requirement. It was shown that the reference case of injected proton beam condition (400 MeV-250 kW and 20 pA/cm2 of beam current density) was applicable to the TEF-T target. Further studies to improve irradiation performance are under way.

TEF-P is a critical assembly, which can accept the 400 MeV-10 W proton beam for the spallation neutron source. The purposes of TEF-P are the experimental validation of the data and method to predict neutronics of the fast subcritical system with spallation neutron source, demonstration of the controllability of a subcritical system driven by an accelerator, and basic research of reactor physics for transmu­tation of MA and LLFP. The distinguishing points of the TEF-P in comparison with existing experimental facilities can be summarized as follows: (1) both the high — energy proton beam and the nuclear fuel are available, (2) the maximum neutron source intensity of about 1012 n/s is strong enough to perform precise measurements even in the deep subcritical state (e. g., keff = 0.90) and is low enough to easily access the assembly after the irradiation, (3) a wide range of pulse width (1 ns-

0. 5 ms) is available by the laser charge exchange technique, (4) MA and LLFP can be used as a shape of foil, sample, and fuel by installing an appropriate shielding and remote handling devices.

Along with the design study of the TEF, R&D for the components required for TEF, such as the laser charge exchange technique to extract a very low power proton beam, test manufacturing of MA fuel-handling devices, and operation of lead-bismuth test loops are under way. From the experimental results of the laser charge exchange technique, beam extraction in the magnetic field is successfully demonstrated. Mockup of the coolant simulator block and remote handling mech­anism for pin-type fuel loading has been done. An effective method to remove polonium with a standard stainless mesh filter was established through the hot experiments. Significant improvement of analysis accuracy of actual ADS was expected by critical experiment with MA fuel at TEF-P.

When the target of TEF-T operates with a full power beam, a fast neutron spectrum field is formed around the target and it is possible to apply multipurpose usage. Various research plans have been proposed, and layout of the experimental hall surrounding the target is under way. Basic physics application such as mea­surements of nuclear reaction data is considered as one of the major purposes. We called for a preliminary letter of intent to encourage the project. Requests for multi­purpose usage will be taken into account in the facility design of TEF.

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.

Experiments for NRD Developments

To evaluate systematic uncertainties in NRD, we have started experimental study at GELINA [5] under the collaboration between JAEA and EC-JRC-IRMM. The items to be studied are as follows: (1) particle size, (2) sample thickness, (3) presence of contaminated materials, (4) sample temperature, and (5) the response of the TOF spectrometer [3, 17]. Some experiments has been performed at GELINA [1821]. A resonance shape analysis code, REFIT [22], has been adopted for the data analysis.

Experiments on sample thickness were carried out at the 25-m TOF neutron beam line of GELINA. Cu plates with various thicknesses were measured with an NRTA method. Peaks at the 579 eV resonance of 63Cu were analyzed with the REFIT program. The evaluated areal densities are compared with the declared values, which were derived from measurements of the weight and the area of the



Fig. 2.3 Ratios of evaluated and declared areal densities. The 579 eV transmission peaks of 63Cu were analyzed with REFIT. Open circles indicate the results analyzed with the resonance param­eter values taken from Mughabghab [23] (#6 in Fig. 2.4), and closed circles represent tentatively introduced values (#7 in Fig. 2.4), which reproduce the areal densities of Cu plates better. The lines are guides for the eye. Note: We also analyzed the transmission spectrum of a 2-cm-thick copper sample with the parameter #6. The obtained fitted curve, however, did not reproduce the peak shape at all. Thus, the misleading open circle data point was removed



Fig. 2.4 Experimentally obtained 579 eV 63Cu resonance parameters. Each data point is taken from different references (#1 [24], #2 [25], #3 [26], #4 [27], #5 [28], and #6 [23]). The data of #6 were utilized by REFIT originally; the data of #7 are tentatively introduced to reproduce the experimental transmission dips samples. Figure 2.3 shows the results. The abscissa is the thickness of the Cu plates and the ordinate is the ratio of evaluated and declared areal densities. The open circles are the results analyzed with the resonance parameter values taken from Mughabghab [23] (#6 of Fig. 2.4); the closed circles are the results analyzed with the tentatively introduced values (#7 of Fig. 2.4), which reproduced the areal density of Cu plates better. Figure 2.4 shows measured 63Cu resonance parameters. The 579-eV resonance parameters of 63Cu may require being reevaluated. It should be emphasized that survey of the total cross sections of Pu and U isotopes is quite important to quantify NM.


We have proposed NRD for measurements of NM in particle-like debris of MF. The NRD system utilizes a compact neutron TOF system equipped with a neutron detector for NRTA and high-energy-resolution and high-S/N у-ray detectors for NRCA/PGA. The rough design of a NRD facility is given. The capacity of NM measurements in the facility has been shown. Experiments on systematical uncer­tainties caused by sample properties, such as sample thickness and uniformity, are in progress under the collaboration between JAEA and EC-JRC-IRMM. The importance of confirmation of nuclear data has been shown in the case of Cu thickness measurements by NRTA.

Acknowledgments The research and development have been carried out under the agreement between JAEA and EURATOM in the field of nuclear materials safeguards research and devel­opment and are supported by the Japanese government, the Ministry of Education, Culture, Sports, Science and Technology in Japan (MEXT).

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