Microfission chambers loaded with highly enriched fissile materials are widely used for measuring power in reactors. The neutron sensitivity of the microfission chamber is a key parameter that determines the accuracy of the power measurement. To evaluate the performance of the FC4A microfission chamber, in this work, we introduced an accurate and validated model of the microfission chamber, a performed Monte Carlo simulation of the neutron sensitivity of the microfission chamber with GEANT4 code, and conducted an irradiation experiment on the neutron irradiation effect platform #3 of the Xi'an Pulsed Reactor. We compared the simulated sensitivity with the experimental results, which showed that the sensitivity obtained from the simulation was in good agreement with the experimental results. In addition, we studied the impact of the design parameters of the fission chamber on the calculated neutron sensitivity of the microfission chamber.

基于MCNP程序建立了西安脉冲堆热中子源设计的蒙特卡罗深穿透耦合屏蔽计算方法;采用MCNP临界源模型计算了热柱方腔前表面的中子 、伽马平面源的参数,并与实验值进行了对比,给出了平面源的修正系数;基于中子、伽马等效平面源,采用新型硼铝复合材料以及铅、铋等材料,优化设计了热中子束流滤束装置,给出了热中子束流滤束装置的升级改造方案,得到热中子通量密度较原设计方案提高3倍、中子伽马通量密度比值大于10的平行热中子束,且束流外侧区域的中子、伽马本底剂量率接近0.025 mSv/h的辐射防护标准.

A neutron flux measurement system based on CVD diamond detector was established, and system response characteristics of neutron and gamma was analyzed. Pulsed neutron flux measurement was carried out under different working mode at Xi'an Pulsed Reactor (XAPR). The signal of diamond detector was coincidence well with that of Pulsed Power Meter in time response, which can show the variation of neutron flux with time. Finally, a comparative analysis of the signal amplitude of diamond detector and pulse peak power of the XAPR was conducted. Wide bandgap semiconductor detector has strong radiation resistance performance and fast pulse response, and its dark current is usually in the order of pA when working in the current mode, the Full Width of Half Maximum response to the repetition frequency of nanosecond pulse hard X-ray source is about 2.2 ns, diamond detector can also be used for neutron flux measurement. Xi'an pulsed reactor (XAPR) has two kinds of operation mode, steady mode and pulse mode. Both of which can carry out electronic components and subsystem steady neutron radiation effect and the transient radiation effect research in the irradiation experiment apparatus. In order to study the transient radiation effect, it is necessary to obtain the radiation parameters such as the neutron flux over time to evaluate the radiation effects of electronic components and subsystem under the pulse radiation condition. In this paper, the fast neutron flux measurement under different pulse conditions was carried out using the single crystal diamond detector working in the current mode, Fig. 1. Fig. 1. Single crystal diamond detector. Radiation response of diamond detector Neutron spectrum of the displacement damage effect experiment apparatus in the XAPR can be measured by multiple foil activation method, the cumulative flux measurement under different pulse conditions is also using the activation method, with the 58 Ni(n, p), 47 Ti(n, p) threshold activation detector. First of all the average cross section can be obtained according to the neutron energy spectrum and the reaction cross section, therefore the cumulative flux can be calculated using the activation rate of the detector activation rate and the average cross section. Neutron gamma ray ratio of the damage displacement effect experiment apparatus in the XAPR is up to 7.7×10 9 n•cm -2 •rad -1 , Fig. 2. Fig. 2. XAPR and experiment apparatus. The diamond detector response to the neutrons and gamma ray has been simulated by Monte-Carlo method based on the neutron energy spectrum, gamma dose rate which has been measured. In the simulation, we need to know the gamma dose rate of the radiation field, and the gamma energy spectrum which cannot be measured of the radiation field. Here, the Monte Carlo method is used to model the reactor and the radiation experimental apparatus. The gamma-ray energy spectrum and dose rate which can be obtained by measurements of the experimental position have been simulated. They can be both used to simulate the response of diamond detector. The crystal size of the diamond detector used in this paper is Ø5 × 0.4 mm. The response of the diamond detector to the neutron and gamma rays of the radiation field needs to be simulated separately by the Monte Carlo method. The interaction between neutron and diamond detector is mainly composed of scattering and absorption with C nucleus, simulating the energy deposition of a series of charged heavy ions, such as C recoil nucleus, proton, α particle, etc. in diamond detection. The diamond detector used for pulsed neutron flux measurement needs to work in current mode, so the sum of energy deposition of each charged particle in the diamond detector should be recorded in the simulation. Gamma rays deposited energy with C atom of diamond detector by photoelectric effect, Compton effect and electron pair effect, and only recorded the average energy deposition. The ratio of neutron and gamma ray in the irradiation apparatus has been determined by measurement. According to this ratio, the average energy of neutron and gamma in diamond detector can be obtained respectively, the result is shown in Table 1. Table 1. Simulate Results

为评估缓发中子探测器监测脉冲堆燃料元件破损的响应特性,建立了MCNP中子源模型和3He球形计数管探测器模型,应用MC方法计算给出了最佳监测位置和该位置处不同聚乙烯厚度下探测器对缓发中子的探测效率曲线.结果表明:最佳监测位置紧贴换热器外表面且偏向入口端,偏移距离和冷却剂贯穿换热器耗时有关,耗时越长,偏移越大;探测效率曲线的变化趋势取决于平均反应截面与中子注量率的积,采用聚乙烯慢化层在提高平均反应截面的同时会降低中子注量率,存在最优厚度4 em使探测器对缓发中子探测效率最大,其值约为1.1×10-5.

Background:Accurate neutron energy spectra and neutron flux parameters are needed to provide the neutron beam service for the reactor users.Purpose:The aim is to accurately measure the fast neutron fluence in Xi'an pulsed reactor.Methods:A neutron fluence monitor method based on 103Rh(n,n')103Rhm reaction was presented.According to the difference of half-life of interference nuclides,cool-down time and measuring time are properly selected to reduce impacts of interference nuclides during the 103Rh(n,n')103Rhm activation rate measurements.On the basis of 152Eu calibration experiment results,the Monte Carlo model of detector was established with optimized detector size by genetic algorithm,the problem of detector efficiency calibration and ray self-absorption was solved.Results:A fast neutron fluence monitor experiment based on this method has been performed,the result of which was consistent with Fe,S,and Ni.The uncertainty of measurement is 13.1％.Conclusion:The establishment of this method successfully solved the problem of fast neutron fluence monitoring in Xi'an pulsed reactor.

The detection efficiency of five kinds of symmetric multi-facet detectors for neutrons with various incident angles was calculated by the Geant 4 code .The calcula-tion results indicate that the efficiency of the detector with symmetric multi-facet struc-ture shows good consistency for neutrons with different incident directions .The spheri-cal proportional detector attached with six circular semiconductor detectors has the most balanced performance in all aspects , thus has the potential to apply in the spectrum measurement of uncollimated neutron field .

The issue of pulse pile-up is frequently encountered when the neutron is detected in intense radiation environment,which will distort the pulse shape and the neutron spectra.In this paper,a digital method of pulse pile-up identification and reconstruction was presented to solve the problem.The pulse pile-up was firstly identified by detecting the downward-going zero-crossings in the first-order derivative of the original signal,and then the constituent pulse was recovered based on four constructed models.The experiment results show that the method is capable of identifying pulse pile-up accurately,as well as disentangling two pulses with a minimum time interval of 20 ns,and it both improves neutrons' effective counting rate and corrects the distorted neutron pulse spectrum due to pile-up event.

The uncertainty analysis is a key link in activation neutron spectrum measure‐ment .According to SAND‐Ⅱ activation neutron spectrum unfolding process ,an uncer‐tainty analysis Monte‐Carlo method of neutron spectrum measurement based on cross‐section covariance ,activation rate and prior spectrum information was presented .First , the sampling method of cross‐section covariance based on linear transformation was built . Then , the error was computed with MCNP , and the uncertainty of prior spectrum was obtained using iteration method .Finally ,combining uncertainty of activa‐tion rate measurement ,the neutron spectrum uncertainty using Monte‐Carlo sampling algorithm was calculated . In the verification of this method with 252 Cf spontaneous fission spectrum ,its uncertainty analysis result was more comprehensive than that of traditional method .The uncertainty analysis of a neutron spectrum measurement for Xi’an Pulsed Reactor was analyzed .The analysis result shows that the method is more conservative .

A simulation approach is developed to obtain the linear energy transfer (LET) spectrum of all secondary ions and predict single event upset (SEU) occurrence induced by neutron in memory devices. Neutron reaction channels, secondary ion species and energy ranges, and LET calculation method are introduced respectively. Experimental results of neutron induced SEU effects on static random access memory (SRAM) and programmable read only memory (EEPROM) are presented to confirm the validity of the simulation results.

给出了一种利用BP神经网络从中子活化测量数据中直接求解1 MeV等效中子注量的新方法.该方法选用了含两层隐藏层的BP神经网络,并围绕先验谱建立输入输出集合,对网络进行训练、检验和测试,最终形成含20个BP神经网络的网络群.该BP神经网络群可实现在输入测量活化率数据后,直接输出相应的1 MeV等效中子注量.利用该方法求解了西安脉冲堆大空间中子辐照实验平台内的1 MeV等效中子注量.与实测中子能谱计算的1MeV等效中子注量结果对比,二者偏差小于6.6％.

The experiments of reactor neutron irradiation which induce dark signal increase in COTS array CCDs are presented. The flux of the reactor neutron beams was about 1.33 × 108 n/cm2s. The three samples were exposed to 1MeV neutron-equivalent fluences of 1 × 1011, 5 × 1011, and 1 × 1012 n/cm2, respectively. The mean dark signal (KD), dark signal non-uniformity (DSNU), and dark signal spikes (hot pixels) versus neutron fluence are investigated. The degradation mechanisms of the dark signal in CCDs are analyzed. The mean dark signal increase due to neutron displacement damage appears to be proportional to displacement damage dose. The dark images from the CCDs irradiated by neutrons are presented to investigate the generation of dark signal spike. The 1D and 2D figures which show the output signal voltage of pixels in dark images irradiated by different neutron beam fluences, are presented to compare the degradation of KD, DSNU, and dark signal spike.

使用ORIGEN2软件计算了西安脉冲堆2MW满功率连续运行和3种断续运行至燃耗末期的堆芯放射性积存量;选取计算结果中最大积存量4.13×1016 Bq,辅以燃料元件包壳全部破损的极端事故假设和保守地释放模型计算放射性物质环境释放量.结果表明,在1 min内的放射性裂变产物释放高达41.0％,有效释放时间大概持续5 min,最大环境释放量为4.54× 1012 Bq.

The large space neutron irradiation platform in Xi ’an Pulsed Reactor was built in 2010 .The designed parameters ,measurement method and experimental result of radiation field of the experiment platform were introduced in the paper .With a prior spectrum calculated by Monte-Carlo method , the neutron spectra of the experiment platform were measured by using multiple-foil activation technique ,and unfolded with genetic algorithm (GA ) . The unfolding result of GA is in accordance well with the result of SAND-Ⅱ method ,and the validity of GA was proved .γ-ray absorbed dose rate was measured by using thermoluminescent dosimeters . The results show that the radiation field parameter of the experiment platform meets the design requirement .

Working principle of the delayed neutron monitor is briefly introduced .Its domestic development is also covered in the paper .Technical essentials on how to discover a failed fuel element using delayed neutron monitor are given in details at the end:including discrimination of short -life isotopes , choice of detector , se-lection of moderator together with the optimization of its thickness , and shielding of outside noise neutrons .

Xe-128 in I-127 transmutation targets irradiated by Xi'an pulse reactor was analyzed by using Quadrupole Mass Spectrometer (QMS) to calculate transmutation rate and release rate. Transmutation is an alternative method to transform long-lived toxic radio nuclides to short-lived radio nuclides or stable nuclides. 127I targets were irradiated in Xi'an Pulse Reactor for the first time to simulate the transmutation behavior of 129I. After irradiation, 128Xe in the target was determinate with online isotope dilution method by Quadrupole Mass Spectrometer (QMS), so that the transmutation rate and release rate could be calculated. Release behaviors of 128Xe were investigated through melting and step-heated modes. The stepwise heating experiment was done by puncturing method. The result shows the release rate in melting mode is 99.9% (RSD = 1%) and increases with the temperature in stepwise heating mode. The technique proposed in this work is suitable for accurate determination of transmutation rate of 129I.

Neutron energy spectrum and fluence of the devices in a reactor are important parameters for the users. In this paper, the thermal neutron spectrum of the thermal column in Xi’an pulse reactor were measurement by the method of time of flight (TOF), and the peak of the energy spectrum is 0.0248±0.0006eV and the neutrons average energy is 0.048±0.001eV. With the spectrum results, the average cross section of 235U(n,f) and 197Au(n,γ) were calculated are respectively 92.08barn and 538.86barn. With the average cross sections, we measured the neutron fluence with the fission chamber and 197Au(n, γ) activation method. The measurements of the two methods are respectively 1.08×109n/m2s and 1.13×109n/m2s. We also calculated the standard uncertainty of the measurements: the energy spectrums’ is less than 5%, and the neutron fluence’ is less than 2.2%.

In this paper,we report the development of a system for 3He and fast neutron sandwich spectroscopy.It consists of a gas-filling system,two planar Si detectors,digital coincidence electronics with three channels and data handling system.Detail descriptions are given on numerical simulations,and experiment calibrations,of the subsystems to optimize the detection efficiency with reduced coincidence backgrounds.The results show that the work is of help for further studies on 3He fast neutron sandwich spectroscopy.

The fuel element burnup measurement theory,the equipment and experimental method are described in this paper.The key techniques for burnup measurement,such as neutron/Gamma self-absorb correction,137Cs activity correction and detect efficiency scale,are emphatically expatiated.The burnup of D5 and G14 are measured and the measurement results are compared with the theoretical calculation.The results show that the result of experimental measurement and theoretical calculation are accordant in uncertainty area,and the system is credible.