Vol.30,

The paper presents GERO (GEneral ReadOut), a general readout ASIC based on a switched capacitor array for micro--pattern gas detectors. It aims at providing general readout electronics for low to medium event-rate gas detectors with high sampling frequency, configurable storage depth, and data digitalization. The first prototype GERO chip integrates 16 channels and was fabricated using a 0.18 µm CMOS process. Each channel consists of a sampling array working in a ping-pong mode, a storage array with a 1024-cell depth, and 32 Wilkinson analog-to-digital converters. The detailed design and test results are presented in the paper.

499

This report presents a design system based on the use of CsI(TI) detectors to search for lost radioactive sources that are dangerous and harmful to individuals, including searching for persons. For this purpose, the GEANT4 simulation toolkit was utilized to develop a system based on thee detectors. Various simulated analyses were performed on the dose rate of the thee detectors using different source–detector distances and detector separation. There was good agreement between the simulated results and the experimentally measured data. A new method was discussed to detect and search for radioactive sources based only on the dose rate in detectors with source activity. Numerical analyses were performed based on the measured dose rates and the difference of distances to determine the actual location of the lost single or multiple γ-ray sources at a specific angle. The detection limit was calculated from the background radiation to establish the sensitivity and capability of the proposed detector system. This system can be applied in fields in which it is necessary to locate unknown radioactive sources.

574

Energy spectra of neutrons are important for identification of unknown neutron sources and for determination of the equivalent dose. Although standard energy spectra of neutrons are available in some situations, e.g., for some radiotherapy treatment machines, they are unknown in other cases, e.g., for photoneutrons created in radiotherapy rooms and neutrons generated in nuclear reactors. In situations where neutron energy spectra need to be determined, unfolding the required neutron energy spectra using the Bonner Sphere Spectrometer (BSS) and Nested Neutron Spectrometer (NNS) have been found promising. However, without any prior knowledge on the spectra, the unfolding process has remained a tedious task. In this work, a standalone numerical tool named "NRUunfold" was developed which could satisfactorily unfold neutron spectra for BSS or NNS, or any other systems using similar detection methodology. A generic and versatile algorithm based on maximum-likelihood expectation-maximization method was developed and benchmarked against the widely used STAY’SL algorithm which was based on the least squares method. The present method could output decent results in the absence of precisely calculated initial guess, although it was also remarked that employment of exceptionally bizarre initial spectra could lead to some unreasonable output spectra. The neutron count rates computed using the manufacturer’s response functions were used for sensitivity studies. The present NRUunfold code could be useful for neutron energy spectrum unfolding for BSS or NNS applications in the absence of a precisely calculated initial guess.

611

Rong Liu，Xing-Yan Liu，Yi-Wei Yang，Jie Wen，Zhong-Wei Wen，Zi-Jie Han，Zhi-Zhou Ren，Qi An，Huai-Yong Bai，Jie Bao，Ping Cao，Qi-Ping Chen，Yong-Hao Chen，Pin-Jing Cheng，ZengQi Cui，RuiRui Fan，Chang-Qing Feng，Min-Hao Gu，Feng-Qin Guo，Chang-Cai Han，Guo-Zhu He，Yong-Cheng He，Yue-Feng He，Han-Xiong Huang，Wei-Ling Huang，Xi-Ru Huang，Xiao-Lu Ji，Xu-Yang Ji，Hao-Yu Jiang，Wei Jiang，Han-Tao Jing，Ling Kang，Ming-Tao Kang，Bo Li，Lun Li，Qiang Li，Xiao Li，Yang Li，Yang Li，Shu-Bin Liu，Guang-Yuan Luan，Ying-Lin Ma，Chang-Jun Ning，Bin-Bin Qi，Jie Ren，Xi-Chao Ruan，Zhao-Hui Song，Hong Sun，Xiao-Yang Sun，Zhi-Jia Sun，Zhi-Xin Tan，Hong-Qing Tang，Jing-Yu Tang，Peng-Cheng Wang，Qi Wang，Tao-Feng Wang，Yan-Feng Wang，Zhao-Hui Wang，Zheng Wang，Qing-Biao Wu，Xiao-Guang Wu，Xuan Wu，Li-Kun Xie，Han Yi，Li Yu，Tao Yu，Yong-Ji Yu，Guo-Hui Zhang，Jing Zhang，Lin-Hao Zhang，Li-Ying Zhang，Qing-Min Zhang，Qi-Wei Zhang，Xian-Peng Zhang，Yu-Liang Zhang，Zhi-Yong Zhang，Ying-Tan Zhao，Liang Zhou，Zu-Ying Zhou，Dan-Yang Zhu，Ke-Jun Zhu，Peng Zhu

To verify the performance of the Neutron TOtal Cross-Section Spectrometer (NTOX), the neutron total cross-section of carbon is initially measured in the energy range of 1 eV to 20 MeV using the time-of-flight method. The measurement is performed at the Back-n White Neutron Source with a 76 m time-of-flight path using the China Spallation Neutron Source. A multi-layer fast fission chamber with ²³⁵U and ²³⁸U is employed as the neutron detector. The diameter and thickness of the natural graphite sample are 70 mm and 40 mm, respectively. Signal waveforms are collected using a data acquisition (DAQ) system. Off-line data processing was used to obtain the neutron time-of-flight spectra and transmissions. The uncertainty of the counting statistics is generally approximately 3% for each bin in the energy range of 1 MeV to 20 MeV. It is determined that the results for the neutron total cross-section of carbon obtained using ²³⁵U cells are in good agreement with the results obtained using ²³⁸U cells within limits of statistical uncertainty. Moreover, the measured total cross-sections show good agreement with the broadening evaluated data.

564

Calibrations were performed for three types of neutron ambient dose equivalent rate meters, i.e., Aloka TPS-451C (Hitachi), KSAR1U.06 (Baltic Scientific Instruments) and Model 12-4 (Ludlum), using a standard field of a ²⁴¹Am-Be source. The measured total neutron ambient dose equivalent rates, H*(10)tot', were analyzed to obtain the direct neutron ambient dose equivalent rates, H*(10)dir', using the ISO 8529-2 recommended the generalized-fit method (GFM), semi-empirical fit method (SEM), and reduced-fitting method (RFM) fit methods. The calibration factor (CF), defined as the ratio between the conventional true value of the neutron ambient dose equivalent rates in a free field, H*(10)FF', and H*(10)dir', was evaluated as one of the important characteristics of the neutron meters in the present work. The fitting results show that the H*(10)dir' values of the meters are in good agreement within the theoretical data within 4%. The averaged CFs of the three neutron meters were evaluated as 0.99 ± 0.01, 1.00 ± 0.03 and 0.99 ± 0.08, respectively. The largest standard uncertainty of these values was determined to be approximately 18.47% (k=1). The standard uncertainty of the CFs obtained using the RFM method was less than 4.23% (k=1), which is the smallest uncertainty among the three methods.

468

A compact interdigital H-mode drift-tube linac (IH-DTL) with the alternating phase focusing (APF) method, working at 325 MHz was designed for an injector of a proton medical accelerator. When fed in with a proper RF (radio frequency) power, the DTL cavity could establish the corresponding electromagnetic (EM) field to accelerate the "proton bunches" from an input energy of 3 MeV to an output energy of 7 MeV successfully, without any additional radial focusing elements. The gap-voltage distribution which was obtained from the CST^® Microwave Studio software simulations of the axial electric field was compared with that from the beam dynamics, and the errors met the requirements within ±5%. In this paper, the RF design procedure and key results of the APF IH-DTL, which include the main RF characteristics of the cavity, frequency sensitivities of the tuners, and coupling factor of the RF power input coupler are presented.

498

Electrorefining of nickel in LiCl-KCl melt was investigated using electrochemical techniques. Nickel products after electrorefining were characterized by X-ray diffraction, X-ray fluorescence, and scanning electron microscopy. Both cyclic voltammetry and square wave voltammetry results suggested that Ni²⁺ was directly reduced to Ni metal in LiCl-KCl. Based on a preliminary study on the electrochemical behavior of nickel and chromium, electrorefining was carried out under constant potential, whereupon, deposits were formed on the cathode. The purity of nickel increased from 72.62% in the original alloy to 99.83% in cathodic deposits, as determined by inductively coupled plasma atomic emission spectroscopy analysis. Almost all the nickel in the alloy could be recovered during the electrochemical process with >90% current efficiency. A lower concentration of NiCl₂ in LiCl-KCl was found to be favorable for nickel electrorefining, as increased NiCl₂ concentration caused severe corrosion of the nickel anode at the gas-liquid interface due to the accumulation of Cl₂ gas.

560

Immobilization devices may be a valuable aid to ensure the improved effectiveness of radiotherapy treatments where constraining the movements of specific anatomical segments is crucial. This need is also present in other situations, specifically, when the superposition of various medical images is required for fine identification and characterization of some pathologies. Because of their structural characteristics, existing head immobilization systems may be claustrophobic and very uncomfortable for patients, during both the modeling and usage stages. Because of this, it is important to minimize all the discomfort related to the mask to alleviate patients’ distress and to simultaneously guarantee and maximize the restraint effectiveness of the mask. In the present work, various head immobilization mask models are proposed based on geometrical information extracted from computerized tomography images and from 3D laser scanning point clouds. These models also consider the corresponding connection to a radiotherapy table, as this connection is easily altered to accommodate various manufacturers’ solutions. A set of materials used in the radiotherapy field is considered to allow assessment of the stiffness and strength of the masks when submitted to typical loadings.

473

Interventional radiology has been beneficial for patients for over 30 years of age with the combination of diagnostic and therapeutic methods. The radiation affecting occupationally exposed workers should be evaluated by means of the energy spectra and flux of X-rays in the treatment room. The present study aims to obtain the energy spectra of interventional procedures and study the capability of some detectors to evaluate the dose in interventional procedures. These measurements were performed by silicon-drift, CdTe, and CdZnTe detectors. The energy spectra were corrected by the energy-response curve of each detector. The energy-response curves of silicon-drift and CdTe detectors provided by the manufacturers specification were used. The energy response of the CdZnTe detector was measured by ¹³³Ba and ¹⁵²Eu γ sources. The experimental data were compared with the simulation results, and their perfect agreement provides a way to correct the energy or dose response, which can be used for the personal dosimeter developed by our group. Moreover, the measured energy spectra can be used in individual radiation protection. The present study shows that the CdZnTe detector is a good candidate detector in interventional procedures.

445

⁵⁶Cu is close to the waiting-point nucleus ⁵⁶Ni and lies on the rapid proton capture (rp) process path in Type I X-ray bursts (XRBs). In this work, we obtained a revised thermonuclear reaction rate of ⁵⁵Ni(p,γ)⁵⁶Cu in the temperature region relevant to XRBs. This rate was recalculated based on the recent experimental level structure in ⁵⁶Cu, the recently measured proton separation energy of S_p = 579.8(7.1) keV, together with shell-model calculation, and the mirror nuclear structure in ⁵⁶Co. The associated uncertainties in the rates were estimated by a Monte-Carlo method. Our revised rate is significantly different from the recent results, which were partially based on experimental results; in addition, we found that a result in a previous work was incorrect. We recommend our revised rate to be incorporated in the future astrophysical network calculations.

549

In order to comply with discharge standards, a gas–solid separator is used to remove solid particles from the Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF) system. As a key component, it directly determines system energy efficiency. However, current gas–solid separators, based on activated carbon adsorption technology, result in high pressure drops and increased maintenance costs. In the present study, a new combined gas–solid separator was developed for the TMSR-SF. Based on a simplified Computational Fluid Dynamics (CFD) model, the gas–solid two-phase flow and the motion trajectory of solid particles were simulated for this new separator using commercial FLUENT software. The flow and separation mechanism for this structure were also been discussed in terms of their velocity effects and pressure field distributions, and then the structure was optimized based on the influence of key structural parameters on pressure and separation efficiency. The results showed that the standard k-ε could be achieved and that the software accurately simulated the new combined separator. In this new combined gas–solid separator, coarse particles are separated in the first-stage using rotating centrifugal motion, and then fine particles are filtered in the second-stage, giving a separation efficiency of up to 96.11%. The optimum blade inclination angle and numbers were calculated to be 45° and four, respectively. It is felt that the combined separator could be of great significance in a wide variety of applications.

529

Small modular thorium-based graphite-moderated molten salt reactors (smTMSRs), which combine the advantages of small modular reactors (SMRs) and molten salt reactors (MSRs), are regarded as a wise development path to speed deployment time. In an smTMSR, low enriched uranium (LEU) and thorium fuel are used in once-through mode, which makes a marked difference in their neutronic properties compared with the case when a conventional molten salt breeder reactor is used. This study investigated the temperature reactivity coefficient (TRC) in an smTMSR, which is mainly affected by the molten salt volume fraction (VF) and the heavy nuclei concentration in the fuel salt (HN). The four-factor formula method and the reaction rate method were used to indicate the reasons for the TRC change, including the fuel density effect, the fuel Doppler effect, and the graphite thermal scattering effect. The results indicate that only the fuel density has a positive effect on the TRC in the undermoderated region. Thermal scattering from both salt and graphite have a significant negative influence on the TRC in the overmoderated region. The maximal effective multiplication factor, which shows the highest fuel utilization, is located at 10% VF and 12 mol% HN, and is still located in the negative TRC region. In addition, on increasing the heavy nuclei amount from 2 mol% HN to 12 mol% HN (VF=10%), the total TRC undergoes an obvious change from -11 pcm/K to -3 pcm/K, which implies that the change in the HN caused by the fuel feed online should be small to avoid potential trouble in the reactivity control scheme.

539

The quantification of ionizing and non-ionizing energy deposition plays a critical role in precision neutron dosimetry and in the separation of the displacement damage effects and ionizing effects induced by neutron radiation on semiconductor devices. In this report, neutrons generated by the newly built China spallation neutron source (CSNS) are simulated by Geant4 in semiconductor material silicon to calculate the ionizing and non-ionizing kerma factors. Furthermore, the integral method is applied to calculate neutron-induced ionizing at the CSNS and non-ionizing kerma factors according to the standard neutron nuclear database and the incident neutron spectrum. In addition, thermoluminescence dosimeters are utilized to measure the ionizing energy deposition and six series of bipolar junction transistors are used to measure the non-ionizing energy deposition based on their neutron damage constants. The calibrated kerma factors that were experimentally measured agreed well with the simulation and integral calculation results. This report describes a complete set of methods and fundamental data for the analysis of neutron-induced radiation effects at the CSNS on silicon-based semiconductor devices.

761

Continuous tracking of bunch charges is the key to maintain stable operations in a storage ring in top-up mode. Recently, a precise bunch-by-bunch beam-current measurement (BCM) system has been developed at the Shanghai Synchrotron Radiation Facility. To avoid the influence of longitudinal oscillation on the amplitudes of the sampling points, a method called two-point equilibrium sampling is introduced. The results, obtained during routine operation time, show that the relative resolution of the measurement of the bunch charges is better than 0.02%. With this high resolution, the new BCM system is able to monitor the bunch-by-bunch beam lifetime. By using the filling pattern information, the Touschek lifetime and the vacuum lifetime can also be calculated. In this paper, the principle of the new method and the experiments is presented in detail.

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