2023年度核データ+PHITS合同研究会/Joint Symposium on Nuclear Data and PHITS in 2023

Nuclear fission serves as the most fundamental physics phenomenon underlying nuclear energy. It is quite a complex process of large-amplitude collective motion of finite quantal systems, which produces a huge energy of around 200 MeV/fission, and fission neutrons, which are used to sustain chain reactions, and over 1,000 kinds of fission products are produced. Understanding nuclear fission is also essential in describing the nucleosynthesis in the cosmos since fission recycling is believed to occur in r-process sites. The mass distribution of fission products in the actinide region is known to be characterized by 2 asymmetric peaks, and it turns to a sharply symmetric shape for nuclei heavier than $^{257}$Fm while it is mildly symmetric for pre-actinide nuclei. This change in the mass distributions between the dominance of the asymmetric and symmetric distributions gives us an important clue to understanding how fission proceeds. Even though there is a long history of research on the fission mechanisms, the essential mechanisms are still a big mystery, and many studies are necessary to understand them from both experimental and theoretical approaches.
On the theoretical side, a large number of macroscopic, macro-microscopic, and microscopic theories have been proposed to describe nuclear fission. As dynamical theories, time-dependent density functional theories have been recognized to be the most advanced method. However, the calculations start outside the barrier in this kind of calculation since trajectories do not come out if these calculations are started from the ground state or the second-minimum of the potential energy surface inside the saddle point. In this approach, therefore, we cannot understand how the system overcomes the barriers after forming a compound nucleus. On the other hand, the Langevin approach, where the nuclear fission is treated as a Brownian motion of nuclear shape degree-of-freedom, can describe the dynamical process of the system starting from the ground state or the second minimum.
The purpose of this study is to elucidate the basic mechanism of nuclear fission, namely, how a compound nucleus overcomes the barrier and how it leads to population of symmetric or asymmetric fission fragments. For this sake, we calculate fission trajectories in the 4-dimensional Langevin model[1] and analyze the trajectories as time-sequential data by using a Long Short-Term Memory (LSTM) method of Recursive Neutral Network (RNN). The time-series data used for training are preprocessed and labeled by time steps, the atomic number Z and the mass number A of a fissioning nuclide, event numbers, values of four collective variables, and corresponding momenta. Then, classification of symmetric or asymmetric fission is performed using multiple all-coupled layers. This presentation will report a progress of this study and preliminary results.

[1] C. Ishizuka, M. D. Usang, F. A. Ivanyuk, J. A. Maruhn, K. Nishio, and S. Chiba, Phys. Rev. C 96, 064616 (2017).

$^{239}$Puの熱中性子核反応における核分裂収率を核反応モデル計算コードCCONE~[1] を用いて計算した。実験データと比較して検証した結果を本発表で行う。

これまでの核データにおける核分裂収率の評価は、主に独立核分裂収率と積算核分裂収率を基にして評価値が決められてきた。しかし核分裂収率は、全運動エネルギー(TKE)や即発中性子、崩壊熱など様々な観測量と関連する物理量であり、それらも考慮した評価値の決定が求められている。JENDL-5~[2] でいくつかの改良が実施されたが、依然として核分裂に伴う観測量との相関は総合的に考慮されていない。この問題を解決するために、我々は核反応モデル計算コードCCONE~[1]を用いた核分裂収率計算システムを開発した。この計算システムでは、核分裂直後の収率分布を５-Gaussianモデルで近似することで、TKEや即発中性子、崩壊熱などの核分裂に伴う観測量を系統的に計算することができる。本発表では、この$^{239}$Puの核分裂収率を計算した結果について紹介する。

References
[1] O. Iwamoto, N. Iwamoto, S. Kunieda, et al., The CCONE Code System and its Application to Nuclear Data Evaluation for Fission and Other Reactions. Nuclear Data Sheets, 131 pp. 259–288 (2016).
[2] O. Iwamoto, N. Iwamoto, S. Kunieda, et al., Japanese evaluated nuclear data library version 5: JENDL-5. Journal of Nuclear Science and Technology, 60, pp.1–60 (2023).

The purpose of this research is to design new radionuclides suitable for brachytherapy sources. Brachytherapy is one of the radiotherapy technique which sealed radionuclides in capsules are inserted into the body. Photons, electrons from a sealed source are usually used. Dose rate can be controlled using a variety of irradiation setup, such as placement and irradiation time to the patients. In some case for the treatment, the biological effects of high-dose-rate irradiation of brachytherapy are superior to those of external beam using medical LINAC. Currently, only a limited number of nuclides such as $^{60}$Co, $^{90}$Sr, $^{106}$Ru, $^{125}$I, $^{137}$Cs, $^{192}$Ir, $^{198}$Au are used in clinical practice. Other radionuclide may also have useful dose effects for the treatment, but not all have been investigated yet. In this study, we simulate the dose properties of radionuclides for brachytherapy based on the AAPM (American association of physicists in medicine) TG-43 method [1,2] using PHITS code [3].

References
[1] M. J. Rivard et al., “Updated of AAPM Task Group No.43 Report: A revised protocol for brachytherapy dose calculations.”, Med. Phys. 31(3)2004.
[2] M. J. Rivard et al., “Supplement2 for the 2004 update of the AAPM Task Group No.43 Report: Joint recommendations by the AAPM and GEC-ESTRO”, Med. Phys. 44(9)2017.
[3] T. Sato, Y. Iwamoto, S. Hashimoto et al., “Features of Particle and Heavy Ion Transport code System (PHITS) version 3.02”, J. Nucl. Sci. Technol. 55(5-6), (2018), pp. 684-690.

Experimental nuclear reaction data are essential for understanding nuclear reaction phenomena, developing nuclear theories and models, and evaluating data for nuclear data libraries. Efficient data mining from the Experimental Nuclear Reaction Database (EXFOR)[1] has a potential for utilization of modern computational analysis techniques to find trends, shortcomings, and hidden patterns in the database, which in turn helps improve our knowledge of nuclear physics. The IAEA Nuclear Data Section (NDS) is entrusted with the responsibility of maintaining and facilitating user-friendly access to this data. To fulfill this mandate, the NDS has developed several services, such as the EXFOR web retrieval system, however, the rapid advances of compute infrastructure and the increasing demand to process nuclear data at scale in the context of ML and AI applications enforces us to adhere the FAIR (Findable, Accessible, Interoperable, Reusable) principles for the service implementation.
To facilitate more advanced method in the nuclear data field, we have developed two EXFOR parsing computer programs (EXFOR Parser) to convert the data in the EXFOR format into the widely adopted JSON format. The converted JSON data are used for further processing to extract individual physical observables and generate tabulated data (x, y, dx, dy) where all units of measurement are standardized. Furthermore, we have developed REST APIs and an open web system for easy access and quick visualizations of these converted datasets.

References
[1] N. Otuka, E. Dupont, V. Semkova, B. Pritychenko, A. Blokhin, M. Aikawa, S. Babykina, M. Bossant, G. Chen, S. Dunaeva et al., Nuclear Data Sheets 120, 272 (2014)
[2] V. Zerkin, B. Pritychenko, Nuclear Instruments and Methods in Physics Research Sec- tion A: Accelerators, Spectrometers, Detectors and Associated Equipment 888, 31 (2018)
[3] M.D. Wilkinson, M. Dumontier, I.J. Aalbersberg, G. Appleton, M. Axton, A. Baak,
N. Blomberg, J.W. Boiten, L.B. da Silva Santos, P.E. Bourne et al., Scientific Data 3,
160018 (2016)

We have developed the radiation dose evaluation system for indoor environments named 3D-ADRES-Indoor, which is especially designed for two applications: the estimation of radioactive source distributions with the machine learning technique and the planning of measures against estimated radioactive sources [1,2]. 3D-ADRES-Indoor mainly uses Particle and Heavy Ion Transport code System (PHITS) [3] for the ambient dose rate calculation required for these applications. In this work, a PHITS simulation technique and a numerical method for the latter application have been developed.
For a better planning of measures against radioactive sources, it is necessary to repeat the simulations with different geometrical models that incorporate various measures. However, it generally takes long computational times to execute whole new PHITS simulations with different geometrical models. Therefore, we have developed a PHITS simulation technique to construct the dose rates with specific models using those obtained from the smaller scale simulations that only account for the difference of models. The point of the technique is the decomposition of dose rates using the “counter” feature and the simulation using the “dump source” feature. While this technique requires the decomposed dose rates and dump data obtained by the normal simulation with a prior model, the computational times of the simulations with the models that incorporate various measures to the prior model are significantly reduced compared to the normal simulations.
We consider four kinds of measures against radioactive sources: the decontamination, removal, relocation of contaminated structures, and shielding. Except for the decontamination, the technique explained above is applied. As for the decontamination, a numerical method to optimize decontamination rate of each radioactive source that achieve target dose rates with the minimum cost for the decontamination. In this method, a constrained optimization problem with the loss function composed of the cost function and the penalty function to achieve target dose rates is solved. To solve this numerical optimization problem, Particle Swarm Optimization (PSO) method has been employed.
Users of 3D-ADRES-Indoor can set conditions of four kinds of measures by GUI operations.
PHITS input required for the present techniques are automatically generated depending on the measures. Users can try more variations of measures with the advantage of the present technique.

References
[1] M. Machida et al., “R&D of Digital Technology on Inverse Estimation of Radioactive Source Distributions and Related Source Countermeasures - R&D Status of Digital Platform including 3D-ADRES-Indoor -”, RIST news No. 69 (2023), pp.2-18.
[2] M. Machida et al., “R&D of Digital Technology on Inverse Estimation of Radioactive Source Distributions and Related Source Countermeasures - R&D Status of Digital Platform including 3D-ADRES-Indoor -”, RIST news No. 68 (2022), pp.3-19.
[3] T. Sato, Y. Iwamoto, S. Hashimoto et al., “Features of Particle and Heavy Ion Transport code System (PHITS) version 3.02”, J. Nucl. Sci. Technol. 55(5-6), (2018), pp. 684-690.

The cross sections of neutron-induced charged-particle emission reactions such as (n,p) and (n,$\alpha$) for many nuclides have not been measured as well as those of the neutron capture reaction. In the present work, new measurement technique for neutron-induced charged-particle emission reactions were developed. The new method uses plastic scintillator added with sample material for measurement. The sample-added scintillator attached on a photomultiplier tube is irradiated with neutrons and charged-particles emitted from neutron-induced reactions are detected at the same time. Scintillators including sample materials were fabricated and the fabricated scintillators were tested in irradiation test experiments conducted with the Accurate Neutron Nuclear Reaction Instrument (ANNRI) of the Japan Proton Accelerator Research Complex (J-PARC). Boron nitride (BN) and lithium fluoride (LiF) were chosen as sample materials to mix with scintillator for the test experiments. The 10B(n,$\alpha$)7Li and 6Li(n,t)4He reactions occur in scintillators added with BN and LiF, respectively. The cross sections of the reactions are high and the Q-values are also high. Thus, charged particles from the reactions are easy to detect and these reactions are good for test. To identify charged particles, the pulse shape discrimination (PSD) was also employed. The pulse shape discrimination technique is based on the property of organic scintillators that the decay constant of light output changes depending on the mass and charge of charged particles. Signals from the photomultiplier tube were fed into the CAEN waveform digitizer V1720 that enables us to process signal onboard for the PSD parameter. In addition to the PSD parameter, the time-of-flight and the pulse heights of events were recorded sequentially. As a result, charged-particles were detected and identified successfully. The present contribution will report the results of the test experiments.

In fluoroscopy, radiation shielding effectively reduces radiation exposure to medical staff [1]. However, it is still unclear how to understand where the scattered radiation comes from and how to properly use radiation shields. The purpose of this study is to clearly visualize the direction of scattered radiation to assist in the optimal use of radiation shields.
The Monte Carlo code PHITS [2] was used to simulate the behavior of scattered radiation under fluoroscopy. The direction vector was obtained by counting the number of photons passing through the plane of each voxel. The voxel space divides the entire fluoroscopy room at regular intervals. The simulations included the x-ray tube, C-arm, water phantom, and couch of the C-arm fluoroscopy system. Scattered photons from the patient were depicted by 3D arrows. Cross sections of the dose distribution were superimposed on the direction vectors. A surgeon model was also included to observe the direction of the scattered rays when the height of the protective plate was adjusted.
The directional vector of the radiation radiating around the patient could be visualized; changing the angle of the C-arm affected the direction and intensity of the radiation. The protective plate effectively shielded the surgeon`s head, especially when placed at a height of 130 cm from the floor, resulting in a 99.1$\%$ dose reduction.

References
[1] ICRP, 2018. Occupational radiological protection in interventional procedures. ICRP Publication 139. Ann. ICRP 47(2).
[2] T. Sato, Y. Iwamoto, S. Hashimoto et al., “Features of Particle and Heavy Ion Transport code System (PHITS) version 3.02”, J. Nucl. Sci. Technol. 55(5-6), (2018), pp. 684-690.

In the case of accidents involving inhalation of radioactive materials, there is a need for rapid evaluation of the amount of transuranium (TRU) nuclides such as Pu ingested in the body by measurement from outside of the body. In the measurement, L X-rays, which have energies of 10 to 30 keV and are emitted by internal conversion electrons due to $\alpha$-decay, are used. In order to measure the L X-ray energy spectrum of TRU, the use of a transition edge sensor (TES)-type microcalorimeter with an energy resolution of less than 100 eV is being considered. It is because each daughter nuclide emits a couple of L X-rays between 10 to 30 keV. In this case, the attenuation of L X-rays in body tissues must be taken into account, and information on the position of deposition in the lungs is important. The deposition positions of α-emitters are estimated by the radiation transport code PHITS with a tetrahedral mesh phantom model published by ICRP and a TES-type microcalorimeter using a Sn absorber. The energy spectra deposited to the absorber are calculated. As a result, it was found that the deposition position in the lung was estimated from the intensity ratio of each L X-ray peak in the energy spectrum.

In Boron Neutron Capture Therapy (BNCT) facilities, concrete of the treatment room is activated by neutrons and dose rate in the treatment room is still high after the end of neutron irradiation. The concrete wall surrounding the room is thick enough to reduce dose rate outside the room to a safety level. The concrete in certain area of the wall is highly radioactive. Estimating the amount of radioactivity depending on the depth from the concrete surface will provide important information for evaluating the safety and economic efficiency of future building demolition. In this study, we estimated the depth distribution of the amount of activation in concrete due to neutrons in a BNCT treatment room, the nuclides that contribute to the dose rate after the end of neutron irradiation, and the dose rate in the treatment room by the Monte Carlo radiation transport code PHITS. The wall behind the water phantom, which simulates a patient, is made of ordinary concrete and boron-containing concrete. The calculation results showed that the amount of concrete activation decreased for the boron-containing concrete, and that the depth up to about 50 cm from the surface made a large contribution to the dose rate in the treatment room.

Photonuclear reaction cross-section data are essential for electron accelerator shielding design and possibly nuclear transmutation. So far, photonuclear cross-sections of various target materials have been evaluated up to a photon energy of 200 MeV within the nuclear data libraries, such as JENDL-5 [1]. Almost all data in JENDL-5 have been evaluated based on the experimental reaction cross-section data. However, the evaluations using to the reaction cross-section data are inadequate to provide all information about the emitted secondary particles, for instance, their energy distributions. Recently, the photoneutron energy spectra on the medium and heavy targets at 13 and 17 MeV photon energies have been measured [2–4]. The 13 and 17 MeV photon beams are nearly monoenergetic and have high intensity. Among the data given in [2–4], $^{209}$Bi is one of monoisotopic elements with the available data at both 13 and 17 MeV photon energies, we evaluated the photonuclear data of this target based on the reaction cross-section data by I. Gheorghe \textit{et al}. [5], and the measured photoneutron energy spectra in [2–3]. The evaluation procedure was conducted on the CCONE code system [6]. The photo-absorption cross-sections were evaluated with the giant dipole resonance (GDR) and quasi-deuteron (QD) models. The emission of photoneutrons is described by the exciton model for the preequilibrium process and the statistical model for the compound process. The photoneutron energy spectra have been compared to the experimental data [2–3] to find the connection between the theoretical reaction models and experiment. An adjustment of the multiplying factor for the single neutron average density in the exciton model was made to improve the energy distribution calculations. For 13 MeV photon energy, this evaluation by CCONE code can reach the measured data well. In contrast, the underestimate to the experimental data is still observed for 17 MeV photon energy when the preequilibrium photoneutrons increase.

References
[1] O. Iwamoto, N. Iwamoto, S. Kunieda, \textit{et al}., "Japanese evaluated nuclear data library version 5: JENDL-5", J. Nucl. Sci. Technol., 60(1), (2023), pp. 1-60.

Shield tunneling methods are widely used to construct large underground tunnels. Although the methods are considered safe, subsidence accidents due to underground cavities created during tunnel excavation have occurred in recent years. To prevent such accidents, it is necessary to detect the cavities and take some measures. Various exploration methods such as ground-penetrating radar have been used to detect such cavities. However, it is difficult to detect cavities deeper than 10 m underground using those conventional methods.
To resolve this issue, we propose an exploration method using the muography technique [1]. Muography is a noninvasive exploration method that utilizes cosmic-ray muons. By measuring muon fluxes at multiple locations underground, information on underground density distribution can be obtained as in a CT scan. Our goal is to develop a disaster prevention system for cave-ins. As a first step, a dedicated prototype muography detector is being developed.
In this presentation, we will report the results of the feasibility test using the prototype detector capable of determining the direction of incoming cosmic rays. For the test measurement, clay bricks were piled above the detector to form a cavity. The muon flux was measured with and without the cavity, and the spatial distribution of transmittance was determined. In the obtained distribution, there was a high-transmittance region corresponding to the cavity position. The size of cavity was estimated from that of the high-transmittance region. A PHITS simulation [2] was then performed incorporating a realistic building structure, and it reproduced the experimental results well.

References
[1] L. Bonechi, R. D’Alessandro, A. Giammanco, Atmospheric muons as an imaging tool, Reviews in Physics, 5,100038 (2020)
[2] T. Sato, Y. Iwamoto, S. Hashimoto et al., Features of Particle and Heavy Ion Transport code System (PHITS) version 3.02, J. Nucl. Sci. Technol. 55, 684-690 (2018).

Fission product yields (FPY) is one of the most important nuclear data, which provide information necessary for applications like burn-up calculation in nuclear reactors, environmental assessment of radioactive waste disposal and evaluation for production of valuable isotopes such as $^{99}$Mo for medical imaging. Traditionally, major nuclear data libraries (such as JENDL, ENDF, JEFF, etc.) have established FPY databases for neutron-induced fission at only thermal energy, 0.5 MeV, and 14 MeV where many experiments have been carried out, and the data in the energy range lacking experimental data are obtained by an interpolation in linear-linear basis, which is not consistent with energy dependence of experimental data for many FPY. Therefore, evaluated data on FPY are still insufficient. However, consistent and systematic evaluations of energy dependence of the FPY have been proven to be quite challenging by both theoretical calculation and experimental measurement. In order to improve this situation, we utilized a two hidden-layer Bayesian neural network (BNN) model with data augmentation technique to train and predict evaluated and experimental fission product yields with high accuracy to obtain energy dependence of mass distributions of the FPY. Similar to the normal deep learning, the JENDL-5 FPY data were divided into 2 groups, 80$\%$ for training and 20$\%$ for validation. Additionally, the training data included several experimental cumulative yields and theoretically calculated values. The number of units in each layer and activation function were selected carefully to reproduce the global and fine structure of the mass distribution data. Additionally, the data augmentation is particularly valuable for enhancing the accuracy of specific nuclides. Finally, the predicted results for energy dependence of mass distribution for $^{232}$Th, $^{233, 235, 238}$U and $^{239,241}$Pu for incident energy ranging from 1MeV to 5MeV in BNN model exhibited certain peak structures at fission product mass numbers A = 134 and A = 140-144, which agreed with enhancement due to the shell and even-odd effects, and the standard I and II asymmetric modes of fission from Brosa model [1]. The capability of our BNN to reproduce the mass distributions including the fine structure is evaluated to be an advancement of the similar approach by a group of Beijing University [2], which aims at description of an overall 2-peak structure of the FPY and the fine structure.

References
[1] Brosa U, Grossmann S, Muller A. Nuclear scission. Physics Reports. 1990;197(4):167–262.
[2] Wang ZA, Pei J, Liu Y, et al. Bayesian evaluation of incomplete fission yields. Physical Review Letters. 2019 09;123.

Nuclear data on carbon ion induced reactions in a wide range of energy are needed for purposes such as improvement radiation protection for space exploration and evaluation systems for secondary exposure on radiation therapy. However, it is reported that there is no measurement data on double differential cross sections of incident high energy $^{12}$C particles between 100 MeV/u and 500 MeV/u. Therefore, there is a need to obtain measurement data of double differential cross sections in high energy regions.
In this study, we measured double differential cross sections of charged particles produced by 100 MeV/u carbon ions on $^{12}$C, $^{27}$Al and $^{59}$Co targets. Obtained data are compared with data previously measured by other researchers and a moving source model. Overall good agreements are shown.

Neutron source term is required for shielding design of accelerator facilities. A Time-of-Flight (TOF) technique is applied to get the source term. When TOF cannot be utilized at accelerator facilities (e.g., Continuous Wave (CW) operation), an unfolding method is useful. However, the validity of unfolding is not completely understood.
One of the accelerator facilities where CW operation will be used is Accelerator-Driven System (ADS) [1]. At the ADS facility designated by JAEA, 1.5-GeV proton beam is provided to a Lead-Bismuth Eutectic (LBE) alloy target. Transmutation of minor actinides is performed by neutrons produced by reactions between incident protons and nuclei in LBE. As with other accelerator facilities, radiation shielding is important for the facility.
According to the shielding design [2], there is a high-dose-rate area at 180$^{\circ}$ attributed to a streaming of a beam duct. Thus, it is required that we study the neutron spectrum at 180$^{\circ}$.
At J-PARC, high-intensity neutron beam at 180$^{\circ}$ is available by the reaction between 3-GeV protons and $^{\rm nat}$Hg. The neutron spectrum at 180$^{\circ}$, which is similar to the spectrum by the reaction between proton and LBE, was already measured with TOF [3].
To evaluate the neutron source term at facilities where TOF cannot be used, confirming the reliability of the unfolding is necessary. Thus, the purpose of this study is to acquire the neutron energy spectrum by the unfolding with the $^{209}$Bi(n,xn) reactions and response functions (JENDL/HE-2007 [4] or TALYS [5]).
In our poster, we present our activation measurement at J-PARC, the unfolding, and the comparison with the TOF-spectrum and calculation results.

References
[1] T. Sugawara, Y. Eguchi, H. Obayashi et al., Conceptual design study of beam window for accelerator-driven system with subcriticality adjustment rod, Nucl. Eng. Des., 331, (2018), pp. 11-23.
[2] H. Iwamoto, S. Meigo, K. Nakano et al., Radiation Shielding Analysis of the Upper Structure of an Accelerator-driven System, JAEA-Research 2021-012, (2022), Japanese.
[3] H. Matsuda, H. Iwamoto, S. Meigo et al., Measurement of thick target neutron yield at 180$^{\circ}$ for a mercury target induced by 3-GeV protons, Nucl. Instrum. Meth. B 483, (2020), pp. 33-40.
[4] Y. Watanabe, K. Kosaka, S. Kunieda et al., Status of JENDL high energy file, J. Korean Phys. Soc., 59 (2011), pp. 1040-1045.
[5] A.J. Koning, S. Hilaire, and M.C. Duijvestijn, TALYS-1.0, Proc. Of International Conference on Nuclear Data for Science and Technology 2007, pp. 211-214.

In equation of state of nuclear matter, constraints on parameters of the symmetry energy $S(\rho)$ are important for understanding of the nuclear many-body system which is related to various astrophysical phenomena. The symmetry energy is essential for the neutron matter ($\delta\sim 1$ where $\delta$ is degree of asymmetry), but it is less certain than the symmetric nuclear matter ($\delta\sim 0$). It is known that there is a linear correlation between the slope parameter and neutron skin thickness $\delta R$ in $^{208}$Pb[1]. $\delta R$ can be written as the difference of the neutron and proton RMS radii. However, the uncertainty of the neutron radius in $^{208}$Pb is still large, while its proton radius is precisely determined by electron scattering.
Proton elastic scattering (PES) is one of the powerful probes in determining the density distributions. In the case of PES, the cross sections at very forward angles which is sensitive to the nuclear radius, are mainly caused by the Coulomb scattering. It is difficult to extract the information of the neutron radius. Therefore, we proposed an experiment of the neutron elastic scattering (NES) to precisely determine the neutron radius in $^{208}$Pb.
Recently, we have performed a measurement of the NES at very forward angles (4, 7 degrees) in $^{208}$Pb and $^{40}$Ca. We have designed a new setup with neutron beams at 63 MeV generated by the $^{7}$Li(p, n)$^{7}$Be reaction. To identify scattered neutrons of NES, Time of Flight (ToF) method and Pulse Shape Discrimination (PSD) method have been applied to the measurement of NES.
We have performed the measurement of angular distribution of NES at $\theta_{c.m.}=4.098,\,4.571,\,6.981,\,7.458$ degrees of $^{208}$Pb(n, el) scattering and $\theta_{c.m.}=4.199,\,4.683,\,7.152,\,7.640$ degrees of $^{40}$Ca(n ,el) scattering. However, angular distribution that we have measured has large statistical errors compared to theoretical requirements. Major factor in the large statistical errors is background neutrons. To distinguish between the background neutrons and the scattered neutrons in $^{208}$Pb and $^{40}$Ca, beam collimation and neutron shielding were essential. Neutron shielding for the experimental setup was calculated by Particle and Heavy Ion Transport code System (PHITS)[2]. In this poster presentation, the details of the experimental setup and feasibility test with PHITS will be discussed.

References
[1] X. Roca-Maza, M. Centelles, X. Viñas et al., “Neutron Skin of $^{208}$Pb, Nuclear Symmetry Energy, and the Parity Radius Experiment”, Phys. Rev. Lett. 106(25), (2011), pp. 252501.
[2] T. Sato, Y. Iwamoto, S. Hashimoto et al., “Features of Particle and Heavy Ion Transport code System (PHITS) version 3.02”, J. Nucl. Sci. Technol. 55(5-6), (2018), pp. 684-690.

The extension of the nuclear fuel life has always been seen as an effective method to improve the economic viability of nuclear reactors. Nonetheless, this has always been hampered by the $^{235}$U 5 wt$\%$ limitation due to criticality concerns [1].An increase in $^{235}$U above the 5 wt$\%$ threshold would mean a major reformulation of both reactor criticality and safety assessments for the present nuclear reactors. The Erbia-credit super high burnup (Er-SHB) fuel is an innovative configuration that allows for the fuel life to be extended while providing several physical improvements (i.e., less downgrade of the flux distribution, improving the intrinsic reactor safety parameters, better control of the transient power phase). Moreover, the negative reactivity introduced by Erbium offers a means to increase the enrichment of $^{235}$U > 5 wt$\%$ while treating the fuel as if the enrichment of $^{235}$U were to be lower than 5 wt$\%$ as in present LWR reactions. Meaning that, this new fuel configuration could be used in present LWR reactors without any major modification to the facilities [1,2]. Nonetheless, for this to be achievable, the accuracy of the nuclear data for the neutron capture cross section of Erbium needs to be improved [3].
The present experiments were performed in the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) at the Materials and Life Science Facility (MLF) of the Japan Proton Accelerator Research Complex (J-PARC) using Li-glass detectors to measure the neutron total cross section; and NaI(Tl) and Ge spectrometers to determine the neutron capture cross section in separate measurements. Several samples of $^{nat}$Er with different thicknesses were measured in the present experiment to improve the accuracy for the cross sections.
In this study, the preliminary results for the $^{nat}$Er neutron total and capture cross sections measured with Li-glass, Ge and NaI(Tl) detectors are presented and compared. For the neutron capture cross section, the preliminary results were obtained relative to the incident neutron flux determined with a boron sample measurement and normalized using an Au sample measurement.

[1] Yamasaki M, Unesaki H, Yamamoto A, et al. Development of erbia-credit super high burnup fuel: Experiments and numerical analyses. Nucl Technol. 2012;177:63–72.
[2] Pergreffi R, Mattioli D, Rocchi F. Neutronics characterization of an erbia fully poisoned PWR assembly by means of the APOLLO2 code. EPJ Nucl Sci Technol. 2017;3:8.
[3] Guglielmelli A. NEA Nuclear Data High Priority Request List [Internet]. Available from: https://www.oecd-nea.org/dbdata/hprl/hprlview.pl?ID=539.

Fast neutron measurements are indispensable technique in the field of experimental nuclear physics and nuclear data measurement. In order to discriminate gamma rays generated by the production of fast neutrons, it is necessary to use a detector capable of discriminating between gamma rays and fast neutrons. As a typical detector for fast neutrons, organic liquid scintillators are widely used. Although liquid scintillators are used in metal containers, there is a problem that the volume decreases over time. Furthermore, import and export procedures are complicated because they are toxic and flammable liquids. On the other hand, plastic scintillators are convenient due to their physical hardness, non-toxicity, and lower flammability. One of the latest pulse-shape discriminating plastic scintillators is EJ-276, but there are few measurements of detector characteristics. This study aims to evaluate the capability to discriminate between neutrons and gamma rays and derive the neutron response function.
In the experiment, the time of flight of neutrons from $^{252}$Cf neutron sources was measured. A $\Phi$5 inch x 2 inch EJ-276 coupled to a photomultiplier tube (PMT) Hamamatsu photonics R1250 and two $\Phi$2 inch x 2 inch EJ-301s coupled to a PMT Hamamatsu photonics R7724 were used. Signals from the PMTs were fed to a digitizer, CAEN-V1730SB, to convert the analog waveforms into digital data. The discrimination capability of EJ-276 and EJ-301 was compared. The response functions obtained from the experimental data were compared with PHITS SCINFUL calculations [1]. The comparison results will be reported in the presentation.

References
[1] D. Satoh, T. Sato, “Improvements in the particle and heavy-ion transport code system (PHITS) for simulating neutron-response functions and detection efficiencies of a liquid organic scintillator” J. Nucl. Sci. Technol. 59(8), (2022), pp. 1047-1060

A variety of unstable nuclear beams with atomic numbers (Z) up to 92 can be produced by the projectile fragmentation and in-flight fission from high intensity U beams at RIBF. Recently, it was found that $^{234−238}$Np can be created by a proton pickup reaction on 1 GeV/nucleon $^{238}$U beam. Owing to the recent developments of the high-Z beams at BigRIPS, energy dependence of the proton pickup reaction on $^{238}$U can be obtained at RIBF. Thus, we conducted an experiment to determine the energy dependence of the production cross section of $^{237}$Np. A test of the production of Np isotopes was performed by using the BigRIPS spectrometer at RIBF in March 2022.
Secondary beams around Z = 90 were produced by a $^{238}$U beam with energies of 345 and 250 MeV/nucleon impinging on a 1-mm-thick $^{9}$Be production target at F0 in BigRIPS.
The particle identification (PID) of the secondary beam was performed using the TOF-Bρ-ΔE method.
To validate the production of the $^{237}$Np$^{91+}$, a two dimensional (2D) Gaussian fitting approach was conducted in accordance with the distribution patterns of neighboring ions of $^{234}$U$^{90+}$, $^{235}$U$^{90+}$, and $^{232}$Pa$^{89+}$. It is found that Np isotope can be counted up with contaminated U/Pa isotopes using the 2D Gaussian fitting technique. The production cross sections of $^{234}$U, $^{235}$U, $^{236}$U, $^{232}$Pa, and $^{233}Pa as well as Np isotopes were derived.
In this presentation, we will report the analysis status of 345 MeV/nucleon.

The control of occupational exposure in fluoroscopy and interventional radiology is critical due to the risk of radiation exposure. Current dosimetry measurements have limitations, such as incomplete whole-body measurements and lack of real-time measurements. To improve radiation awareness, we have developed a system that displays 2D scattered radiation distribution and estimates the surgeon's radiation dose in real time.
Using the Monte Carlo code PHITS [1], a scattered radiation simulation for X-ray fluoroscopy during the examination was performed. Three-dimensional data of scattered radiation was mapped into the X-ray room using an AR marker. A two-dimensional scattered radiation display was created. A real-time scattered radiation exposure system has been developed to track a surgeon's body using AzureKinect. The system's accuracy in distance and dose was tested, comparing measurements with a laser rangefinder and dosimeters.
The system estimated the dose and accurately visualized the radiation distribution. Distance accuracy improved as the surgeon moved closer to the camera. Distance accuracy decreased as the distance from the camera exceeded the body tracking capability. Dose estimates were 0.6 to 1.2 times higher than actual measurements. Dose accuracy was lower in the chest and pelvis areas, likely due to surface mounting of the dosimeters and body tracking of the system for torso measurements.

References
[1] Sato T, Iwamoto Y, Hashimoto S, et al. Features of Particle and Heavy Ion Transport code System (PHITS). version 3.02. J Nucl Sci Technol; 2018; 55

We have measured a small-angle neutron and X-ray scattering (SANS and SAXS) of cement paste to investigate a nanoscale structure of cement paste [1, 2]. Through the in-situ SAXS measurements of cement paste, in particular, we have focused on a fine nanostructure that emerged with time as a shoulder on the SAXS profiles at the high-q region of around 3 nm$^{-1}$. Based on a microstructure model of cement paste [3] and a previous SANS work [4], it is expected that the fine nanostructure consists of calcium silicate hydrate (C-S-H) gel and pore water, where C, S, and H stand for respectively CaO, SiO$_{2}$, and H$_{2}$O in a conventional notation of cement chemistry. The C-S-H is a major hydrate among the cement hydrates and relates closely with the compressive strength of hardened cement paste (HCP).
Recently, to obtain information of the elemental composition of the fine nanostructure, SANS measurements of HCP samples were conducted using a contrast variation method in BL15 TAIKAN of MLF at J-PARC. In addition, the neutron transmissions of the saturated and dried HCP samples were also measured because water contents including these samples were evaluated for subtracting the background due to incoherent scattering of hydrogen from the SANS profiles. In the data analysis for the water contents evaluation, the neutron transmissions which were calculated using the PHITS code were compared with the measured neutron transmissions, where the JENDL-5 ACE library (ACE-j50; neutron induced nuclear data and thermal scattering law data for hydrogen and deuterium in water) [5] was applied to the PHITS calculations. The data analysis of the neutron transmissions is presented together with the data of SANS measurements.

References
[1] K.Y. Hara, M. Ohnuma, Y. Yoda, and K. Hiroi, AESJ 2023 Fall Meeting, 2I18 (2023 Sep 6-8) [in Japanese].
[2] K.Y. Hara, Y. Morinaga, Y. Yoda, and M. Ohnuma, Proc. 2021 Symp. Nuclear Data (2022) pp.109-114 (jaea-conf-2022-001).
[3] M.B. Pinson, \textit{et al}., Physical Review Applied 3 (2015) 064009.
[4] A.J. Allen, J.J. Thomas, and H.M. Jennings, Nature Material 6 (2007) 311.
[5] Web page of JAEA reactor physics team (https://rpg.jaea.go.jp/main/en/ACE-J50/).

The Intranuclear Cascade model have been improved for calculation of alpha incident reaction. The fragmentation reaction which is dominant in alpha incident reactions is calculated using the model that incident alpha particle is broke up according to the probability, which show cluster state in the alpha particle, and density distribution of targets. In addition, the direct knockout reaction is adopted to explain composite particles in low energy region. The calculation results were compared with experimental data of double differential cross sections of charged particles produced from the reaction of alpha particles at incident energy of 230 MeV/u on $^{12}$C, $^{27}$Al, and $^{59}$Co. As a result, good agreements are obtained.

Neutron activation analysis is used in a variety of many fields, such as geology, medicine, and archaeology. In the neutron activation method, a sample is irradiated with neutrons and activated via neutron-induced reactions. The elemental composition is determined from the radioactivity of the sample. In measurement of radioactivities, -rays are often detected with a germanium detector which provides high enough energy resolution to identify $\gamma$-rays from the sample. If the resultant radioactive nuclides emit only $\beta$-rays, $\beta$-rays are detected with a $\beta$-ray detector such as gas counter for high energy $\beta$-rays or liquid scintillator for low energy $\beta$-rays. $\beta$-rays have shorter range than $\gamma$-rays. This property requires the sample to be thin enough for $\beta$-rays to emerge from the sample or to be dissolved into solvent of liquid scintillator. Thus, sample preparation including chemical processing is usually necessary. Measurement of pure $\beta$ nuclides is not as easy as that of $\gamma$-ray emitting radionuclides. To make neutron activation analysis for pure $\beta$ nuclides easier, a new technique is being developed in the present research. In the new method, sample material for analysis is added to plastic scintillator and the scintillator including the sample is irradiated with neutrons. The irradiated scintillator is attached to a photomultiplier tube to count $\beta$-rays to determine the radioactivity. This technique does not require complicated chemical process before or after irradiation. To study the feasibility of this technique, the method to fabricate scintillator including sample material was developed and test irradiation experiments were carried out at the Tokyo Institute of Technology (Tokyo Tech). A gold foil was irradiated with neutrons from the $^{7}$Li(p,n)$^{7}$Be induced by bombarding a lithium target with a proton beam from a Pelletron accelerator of Tokyo Tech. $\beta$-rays from the gold foil were detected with the fabricated scintillator. This contribution will report the current progress and results of the development.

Purpose: At energies above the ($\gamma$,n) threshold, photons can interact with the nuclei of high-Z materials, liberating fast neutrons. The aim of this study was to validate the Monte Carlo simulation with the PHITS code for the TrueBeam LINAC using mono-energy. Additionally, we examined the photon neutron dose surrounding the TrueBeam LINAC's head and investigated the influence of field size on the distribution of photon neutron ambient dose.
Method and Materials: Research group used PHITS codes version 3.29, Japanese Evaluated Nuclear Data Library (JENDL-5.0), and the training data of Varian to simulate the head of TrueBeam LINAC 10 MV. The simulated Percentage Depth Dose (PDD) (Field size 10 x 10 cm$^{2}$, Source Surface Distance (SSD) 100 cm), and simulated crossline at 5, 10, 20 cm depths were compared with the measured data. And then research group used these PHITS codes to calculate photon and neutron dose at twenty-five points around the head of TrueBeam LINAC 10 MV with both two field size 20 x 20, and 0.5 x 0.5 cm$^{2}$.
In measurement: PDD and crossline were measured with Blue Phantom, CC13S ion chamber, TrueBeam LINAC 10 MV photon; Photon and neutron dose at each point of twenty-five points around the head’s TrueBeam LINAC were measured with three Radio-photoluminescence for photon dose and three CR-39 detectors for neutron dose, when TrueBeam LINAC radiated 50 Gy in both field size 20 x 20, and 0.5 x 0.5 cm$^{2}$.
Results: The measured neutron doses were in the range 0.4 – 12.53 mSv (0.5 x 0.5 cm$^{2}$), the range 0.43 – 12 mSv (20 x 20 cm$^{2}$); The measured photon doses were in the range 0.63 – 177.00 mSv (0.5 x 0.5 cm$^{2}$), and 2.23 – 183.33 mSv (20 x 20 cm$^{2}$). The simulated neutron doses were in the range 0.11 - 26.65 mSv (0.5 x 0.5 cm$^{2}$), the range 0.06 – 14.36 mSv (20 x 20 cm$^{2}$); The simulated photon doses were in the range 0.16 – 182.63 mSv (0.5 x 0.5 cm$^{2}$), and 1.58 – 178.84 mSv (20 x 20 cm$^{2}$).
Conclusion: Measured and simulated photon neutron dose showed larger field size increased photon ambient dose distribution, and decreased neutron ambient dose distribution. In vice, smaller field size decreased photon ambient dose distribution and increased neutron ambient dose distribution.