Benchmarking of JEFF-3.2, FENDL-3.0 and TENDL-2014 evaluated data for tungsten with 14.8 MeV neutrons

Integral experiments on tungsten slab samples were carried out on the D-T neutron source facility at China Institute of Atomic Energy.Leakage neutron spectra from the irradiated tungsten target were measured by the time-of-flight technique.Accuracy of the nuclear data for tungsten was examined by comparing the measured neutron spectra with the leakage neutron spectra simulated using the MCNP-4C code with evaluated nuclear data of the JEFF-3.2,FENDL-3.0 and TENDL-2014 libraries.The results show that the calculations with JEFF-3.2 agree well with the measurements in the whole energy range and all angles,whereas the spectra calculated with FENDL-3.0 and TENDL-2014 have some discrepancies with the experimental data.     

Benchmarking of JEFF-3.2, FENDL-3.0 and TENDL-2014 evaluated data for tungsten with 14.8 MeV neutrons

Integral experiments on tungsten slab samples were carried out on the D-T neutron source facility at China Institute of Atomic Energy.Leakage neutron spectra from the irradiated tungsten target were measured by the time-of-flight technique.Accuracy of the nuclear data for tungsten was examined by comparing the measured neutron spectra with the leakage neutron spectra simulated using the MCNP-4C code with evaluated nuclear data of the JEFF-3.2,FENDL-3.0 and TENDL-2014 libraries.The results show that the calculations with JEFF-3.2 agree well with the measurements in the whole energy range and all angles,whereas the spectra calculated with FENDL-3.0 and TENDL-2014 have some discrepancies with the experimental data.     

准直D-T中子束穿过大块板状聚乙烯样品积分实验

本文介绍了中国原子能科学研究院建立的准直中子束积分实验装置。该装置利用T(d,n)⁴He反应产生14.8 MeV脉冲中子束,经1.1 m厚重水泥屏蔽墙上的准直孔道后与样品作用,用飞行时间法测量样品不同方向的泄漏中子谱。首次测量了样品厚度分别为4.5、9、18和27 cm的大块板状聚乙烯样品在30°和50°方向的泄漏中子谱;考虑靶结构、源中子能谱和角分布、脉冲束宽度及探测器效率,利用MCNP程序模拟计算了相同实验条件下的泄漏中子飞行时间谱。实验结果与模拟结果符合较好。     

国产Am-Be中子源4.438MeVγ射线与中子强度比值测量

本工作涉及准确测量国产Am-Be中子源发射的4.438MeVγ射线与中子强度比值R＝S_γ/S_n的实验方法。中子源的中子发射率用锰浴法进行比对测量。用Φ75mm×75mmNaI（Tl）探测器测量中子源的γ能谱;用MCNP程序模拟计算中子引起的γ本底和探头的源峰探测效率。实验与理论计算得到的R值符合得很好。综合评价已发表的R实验值,给出了R推荐值为0.575（1±4.8%）。结果表明,R值可认为是Am-Be源的一标志性特征量。     

A Two-Dimensional Benchmark Experiment for Neutron Transport in Water

For the purpose of providing standard data for checking two-dimensional neutron penetration calculations, fast neutron spectra as well as thermal and epithermal neutron fluxes were measured over a two-dimensional (R, Z) space in water shield using an activation method. Threshold reaction rates were converted to fast neutron scalar flux spectra with the aid of the SAND-II code. These results agree within a factor of 2 with the calculations by a two-dimensional discrete ordinates code PALLAS-2D. Thermal and epithermal neutron fluxes obtained with the Westcott's method agree quite well with the calculated values by the PALLAS-2D code in which the diffusion equation was adopted for dealing with low energy neutrons to reduce the computing time. All experimental results are given in the absolute values.     

AB-BNCT中子靶物理设计分析

质子加速器适用于为硼中子俘获治疗提供中子源,其中子源强及能谱较反应堆中子源更具可调性。中子靶物理计算分析是加速器中子源设计的基础,为其提供粒子能量、流强等参数需求分析,并为靶体结构尺寸设计、中子慢化和屏蔽分析等提供前端参数。本文利用MCNPX蒙特卡罗程序,通过对质子打靶的中子产额和能谱、靶体能量沉积、打靶后靶材放射性活度和中子出射空间角分布等进行研究,提出能量2.5 MeV质子轰击100～200 μm锂靶的设计,并用模拟计算数据论证其合理性。该设计中子源在1 mA流强质子轰击下,源强可达9.74×10¹¹ s^-1;拟设计15 mA、2.5 MeV质子束产生的中子源,在治疗过程中靶材放射性活度累积最大值约为1.44×10¹³ Bq。     

Thermal Neutron Spectra in Light Water Poisoned with Cadmium and/or Indium

The thermal neutron spectra in light water of slab geometry poisoned with Cd and/or In were measured by the time of flight method, using a 20-MeV electron linear accelerator. The thermal spectra were simulated to those in the Pu built-up core of a commercial light water reactor corresponding to a fuel burnup of about 15,000 MWD/T. The results of measurements were compared with calculations based on the S ₄ method using the Haywood scattering law. Fairly good agreement was obtained between the calculated and measured results except in a limited range of energy above the 0.176 eV resonance of Cd. It is concluded that the P ₁ components of the source neutrons as well as the neutron scattering kernel play a significant role in the calculation of the thermal neutron spectra with large flux gradients, and that the scattering kernel of light water based on the Haywood model will be accurate enough to evaluate the infinite multiplication constant k_∞ of light water reactor cores with high fuel burnup within an error of about — 0.17%, as estimated from the uncertainty in the spectrum calculation in the region above the Cd resonance. It is also emphasised from the two- dimensional S ₄ calculations that the effect of reentrant hole perturbation should be evaluated quantitatively in the interpretation of the measured angular neutron spectra produced within finite media.     

油砂地层中子-伽玛能谱Monte-Carlo模拟

本工作应用自行研制开发的中子 伽玛能谱测井的中子 伽玛射线随空间、能量、时间分布的蒙特卡罗模拟软件包计算了裸眼井、井眼里无下井仪器、油饱和砂岩地层、井眼注油条件下沿井轴和井壁的中子 伽玛射线随能量、时间分布及沿井轴和沿井壁的中子 伽玛射线能谱 ,研究了源距对中子 伽玛射线随能量、时间分布以及井轴与井壁中子 伽玛射线能谱的影响。     

T(d,n)4He强流中子发生器的中子能谱和角分布模拟计算

采用将厚靶分割成薄靶的方法对厚氚钛靶、260keV氘束流能量条件下T(d,n)⁴He反应中子源的能谱和角分布进行计算。以分割法计算得到的能谱和角分布数据为基础,建立了D-T中子源Monte-Carlo模拟抽样模型,在考虑中子发生器各元件材料及实验大厅墙壁对快中子的慢化、散射和吸收的条件下,采用MCNP程序对兰州大学3×10¹²s^-1强流中子发生器260keV氘束流能量下的中子能谱和角分布进行了模拟,给出了模拟结果。为检验模拟结果的可靠性,与实验测量能谱进行了比较,Monte-Carlo模拟谱和实验测量谱基本符合。     

Cross Section Assessment by Measurement of Time Dependent Neutron Spectra from Lithium Assemblies

Time dependent neutron spectra from lithium assemblies were measured to assess the neutron cross sections of ⁷Li in ENDF/B-IV, which is important nuclide for the D-T fusion reactor blanket material. Pulsed neutrons produced by D-D or D-T reaction were used to measure leakage neutron spectra from cubical lithium assemblies as a function of time by the use of NE213 liquid scintillator. Calculations of time dependent neutron spectra were carried out by the Monte Carlo code SIMON, which was prepared for this study. The group constants used in these calculations were processed from ENDF/B-IV data. The calculated and the measured neutron spectra were compared for the following three; a stationary spectrum, spectra at each time interval and decay curves for specified energy groups. Discrepancies between the measured and the calculated neutron spectra were found in these comparisons. In order to assure the cause of these discrepancies, some calculations were carried out with recently measured cross sections of inelastic scattering which excite 0.478 and 4.63 MeV level of ⁷Li. It was concluded that some of the neutron cross section data of ⁷Li in ENDF/B-IV should be ameliorated.     

A function to provide neutron spectrum produced from the 9Be + p reaction with protons of energy below 12 MeV

A function to give the total neutron production cross section, angular distribution, and energy spectrum via the ⁹Be + p reaction has been created by fitting experimental data to characterize compact neutron sources with thick Be targets bombarded by protons with energy below 12 MeV. To examine the suitability of the function, calculations of the angle-dependent neutron energy spectra produced in thick Be targets with 4- and 12-MeV protons using the function were compared with corresponding experiments and calculations using the nuclear data libraries of ENDF/B-VII.0 and JENDL4.0/HE. The function was in better agreement with the experiments than the calculations using the libraries except for at backward angles. The ¹¹⁵In(n,n’)^115mIn reaction rates calculated using GEANT4 with source neutrons given by both the function and ENDF/B-VII.0 were compared with that measured at the RIKEN Accelerator-Driven Compact Neutron Source to evaluate the neutron spectrum above 1 MeV. The function slightly overestimated the measurement by 14% and the calculation with ENDF/B-VII.0 underestimated by 35%. It was concluded that the function can be applied in compact neutron source designs.     

Response Functions of Phoswich-Type Neutron Detector for High-Energy Cosmic Ray Neutron Measurement

A phoswich-type neutron detector was developed in order to measure high-energy cosmic ray neutron spectra in aircraft. The neutron detector consists of an EJ309 organic liquid scintillator that is 121.7mm in diameter and 121.7mm in length and is covered with a 15mm thick EJ299-13 outer plastic scintillator. The neutron response functions of the detector are required for the unfolding method to obtain the energy spectrum. The neutron response functions were created based on MCNPX simulations using an anticoincidence mode with the experimental light-output correlations with particle energies, uniformity of light collection and energy resolutions. The light-output correlation with particle energy, the uniformity of light collection and the energy resolutions were evaluated based on experiments. Measurements of neutron response functions were performed using four quasi-monoenergetic neutron beams from 40 to 80 MeV p-Li reactions to verify the calculated results. The calculated response functions show good agreement with the measurements. The angular response of the phoswich detector was confirmed to be isotropic from the calculation. The photon response functions of the detector were also calculated and agreed well with the measurements for 6.129MeV photons. Neutron and photon response matrices were created up to 300 and 50 MeV, respectively, over a wide energy range for experimental flights.     

Inelastic neutron scattering and prompt fission neutron spectra for Np

The secondary neutron spectra (inelastic, elastic, fission) for ²³⁷Np were measured by the neutron time of flight spectrometer of the IPPE at the incident energy range 1–2.5 MeV. The solid tritium target was used as a neutron source. The neptunium oxide (189 g) packed in the low mass stainless steel container was used as a scattering sample. The neutron background due to scattering on the target environment and tritium into the target backing was measured and was calculated with the appropriate model of the neutron source. The data were corrected for neutron background, the scattering on the oxygen and iron nuclei, and the effect of the finite sample size. The fission neutron spectra were measured, evaluated and subtracted from the emission neutron spectra to estimate inelastic neutron spectra and cross-sections. The experimental results were compared with ENDF/B-VI, BROND-2, JENDL-3 neutron data libraries.     

Calculation of neutron and gamma-ray energy spectra for fusion reactor shield design: Comparison with experiment-II

Measured and calculated neutron and gamma-ray energy spectra resulting from the transport of 14 MeV neutrons through a 0.30-m-thick lithium hydride slab and through a 0.05-m-thick lead slab followed by 0.30 m of lithium hydride are compared. Also reported are comparisons of the measured and calculated neutron energy spectra behind an 0.80-m-thick assembly comprised of stainless steel type 304 and borated polyethylene. The spatial dependence of the gamma-ray energy deposition rate measured using thermoluminescent detectors is compared with calculated data. The calculated data obtained using two-dimensional radiation transport methods and ENDF/B-IV cross-section data are in good agreement for all of the experimental configurations. Calculated integral neutron energy spectra agree with the measured data within 5 to 20% depending on neutron energy for the LiH and Pb plus LiH assemblies. The gamma-ray spectra agree within 20% for these slabs. The measured and calculated neutron energy spectra behind the SS-304-borated polyethylene assembly agree within 5% except at neutron energies below 5 MeV where background radiation influences the measured spectra. The gamma-ray energy deposition rates as a function of depth agree within a factor of two at all detector locations.     

70MeV质子加速器产生的白光中子源强度和能谱的计算

用７０ＭｅＶ质子束轰击厚靶可产生白光中子源。选用金属钨作厚靶材料，应用适用于几十ＭｅＶ能区的核反应模型程序计算了中子能谱和角分布。结果表明：７０ｍｅＶ质子束在钨厚靶中终止前有４．６％的质子发生了反应，２００μＡ的７０ＭｅＶ质子束流产生的中子总强度为１．０１×１０１４Ｓ－１，中子平均能量为４．２ＭｅＶ，１０ＭｅＶ以上的中子强度为１．１５×１０１３Ｓ－１，高能中子主要在朝前的小角度区发射。     

Comparison of measured and calculated neutron and gamma ray energy spectra behind an in-line shielded duct

Integral experiments that measure the transport of 14 MeV neutrons through a 0.30-m-diameter duct having a length-to-diameter ratio of 2.83 that is partially plugged with a 0.15 m diameter, 0.51 m long shield comprised of alternating layers of stainless steel type 304 and borated polyethylene have been carried out at the Oak Ridge National Laboratory. Measured and calculated neutron and gamma ray energy spectra are compared at several locations relative to the mouth of the duct. The measured spectra were obtained using an NE-213 liquid scintillator detector with pulse shape discrimination methods used to simultaneously resolve neutron and gamma ray events. The calculated spectra were obtained using a computer code network that incorporates two radiation transport methods: discrete ordinates (with P₃ multigroup cross sections) and Monte Carlo (with continuous point cross sections). The two radiation transport methods are required to account for neutrons that singly scatter from the duct to the detectors. The calculated and measured neutron energy spectra above 850 keV agree within 5–50% depending on detector location and neutron energy. The calculated and measured gamma ray energy spectra above 750 keV are also in favorable agreement, 5–50%, depending on detector location and gamma ray energy.     

Method of Measuring Boron-10 Depletion in Control Blade

The neutron capture cross sections and capture γ-ray spectra of ^143,145,146Nd were measured in the neutron energy region of 10 to 90 keV and at 550 keV. A neutron time-of-flight method was adopted with a 1.5-ns pulsed neutron source by the ⁷Li(p, n)⁷Be reaction and with a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique was applied to observed capture γ-ray pulse-height spectra to derive capture yields. The capture cross sections were obtained with the error of about 5% by using the standard capture cross sections of ¹⁹⁷Au. The evaluated values of JENDL-3.2 and previous measurements were compared with the present results. The capture γ-ray spectra were obtained by unfolding the observed capture γ-ray pulse-height spectra. An anomalous shoulder was observed around 2 MeV in the γ-ray spectra of ^145,146Nd, and the energy position of the shoulder was consistent with the systematics obtained in our previous work.     

New 6Li Sandwich Counter for Measurement of Neutron Spectra in Fast Neutron Systems

A new form of ⁶Li sandwich counter has been developed for use in measurements of neutron spectra in fast critical assemblies. The counters have two large surface area diodes approached very close to each other, on one of which is evaporated the target ⁶Li. This arrangement proves to provide a ten-fold increase of detection efficiency, while the resolution is improved by approximately ½ compared with currently used counters. The new counters were first tested on mono-energetic neutrons, after which neutron spectrum measurements were performed on natural uranium blocks, in the fast source reactor “YAYOI”, in the FCA IV-l-P and in the FCA V-3 core. The resulting spectra are compared with the measurements made with other types of counter and with calculated data. Proof was obtained that the new counter can serve practically and usefully for in-core measurements of fast neutron energy spectrum between 0.8 and 5 MeV.     

加速器中子源大厅内散射中子分布的模拟

本文针对加速器中子源可在较宽能量区间产生单能中子的特点,采用MCNP5对0.2～20 MeV的源中子在加速器中子源大厅内的散射情况进行模拟计算和分析。结果表明,直射中子通量随离源距离的增大呈平方反比衰减,散射中子通量则随离源距离的增大而几乎保持不变;大厅内的散射中子主要来自墙壁的贡献,离墙壁越近散射率越高。能量为0.4 MeV和1 MeV的源中子散射率最高,10 MeV和15 MeV的源中子散射率最低。用中子的宏观散射截面可较好解释散射率模拟结果,中子的弹性散射截面远大于非弹性散射截面,因此弹性散射起主导作用。中子能量大于1 MeV后,散射截面随中子能量增加而减小直至进入一段坪区,散射率也随之降低并进入坪区。结合待测位置处直射、散射中子通量和不同能量的散射中子份额的计算,能解释能量较高的源中子散射率较低的现象。通过在墙壁表面附上一层中子慢化吸收材料的方法可有效减弱中子散射,如5 cm的含硼聚乙烯（10%B4C）可降低散射率约40%。     

Experiments and Calculations of 14 MeV Neutron Streaming through Multi-Layered Slit Assembly

A streaming experiment using a D-T neutron source was carried out to verify the calculational technique for neutron transport in a shield assembly with multi-layered slits. Reaction rate distributions of a small spherical NE213 scintillation detector to fast neutrons were measured in the slits made of 304SS and in the mortar surrounding the slits. The energy spectrum of fast neutrons in the slit was also measured with the same detector. These measurement were compared with calculations using the continuous energy Monte Carlo transport code MCNP. The calculated reaction rates in the slits agreed with the measured ones within experimental and calculational errors. Besides, it is suggested that the attenuation of fast neutrons in the mortar is significantly different from that in the slits and the behavior is nearly traced by the calculation with the MCNP code. The measured and calculated spectra at a position close to the exit inside the lower slit agreed within the both errors.