医院中子照射器I型堆堆芯热工水力分析

针对医院中子照射器Ⅰ型堆(IHNI-1)的堆芯特点和运行工况,建立了适用于IHNI-1反应堆堆芯的热工分析模型,并对模型进行了验证.利用所建模型,计算了 IHNI-1反应堆堆芯热工参数.最后分析了IHNI-1反应堆堆芯入口流量对堆芯出口温度的影响,同时给出了堆芯发生过冷沸腾时的功率计算结果.     

分群结构对IHNI-1堆动态参数计算结果的影响研究

基于WIMS和CITATION程序的计算结果,编制了动态参数计算程序CKPWC(calculating kinetic parameters based on WIMS and CITATION),对医院中子照射器Ⅰ型堆(in-hospital neutron irradiator mark 1 reactor,IHNI-1)的动态参数计算进行了研究.首先使用WIMS计算出均匀化栅元截面以及69群通量,再使用CITATION进行四群扩散计算,最后编制动态参数计算程序,计算了IHNI-1动态参数(缓发中子有效份βeff和中子代时间Λ).经过比较研究发现,分群结构对动态参数的计算结果有很大的影响.给出了计算IHNI-1缓发中子份额和中子代时间的最佳四群分群结构.使用文章中的最佳分群结构思想对西安脉冲堆动态参数进行了验证计算,计算结果与设计值符合一致,说明给出的IHNI-1动态参数计算结果具有一定的可信性.     

事故工况下医院中子照射器I型堆瞬态特性研究

采用RELAPS/SCDAP/MOD3.4程序对医院中子照射器Ⅰ型堆(IHNI-1)在事故工况下的瞬态特性进行研究,对意外大反应性引入和池水丧失事故工况进行了计算和分析,计算结果表明:IHNI-1堆具有良好的固有安全性,在发生大反应性引入和池水丧失事故时,最终能够稳定在较低功率,确保反应堆安全.     

基于MCNP和ORIGEN2耦合程序的IHNI-1型堆裂变产物中毒及燃耗分析

为了准确地计算反应堆的裂变产物中毒和燃耗问题,开发了一套蒙特卡罗方法程序系统.利用通用的燃耗计算方法,基于MCNP和ORIGEN2,编写了相关的数据转换、截面修正、数据接口程序,实现了MCNP和ORIGEN2程序的耦合.采用堆芯精细结构划分,对医院中子照射器Ⅰ型堆裂变产物中毒和燃耗进行了计算分析.     

医院中子照射器I型堆热中子束流孔道等效平面源的模拟计算

采用蒙特卡罗程序MCNP模拟计算了医院中子照射器Ⅰ型堆(IHNI-1)热中子束流孔道出口处的等效平面源.对B堆芯进行了临界搜索计算,模拟计算了热中子束流孔道及出口处中子、γ的束流参数,应用等效平面源模型建立了BNCT等效中子、γ平面源.为人体头颅等效模型剂量分布的快速计算提供了较为可靠的平面源.     

船用反应堆堆芯时空中子动力学仿真软件的研制

根据船用反应堆结构特点与运行方式，建立堆芯三雏两群时空中子动力学仿真模型，研制了船用反应堆堆芯时空中子动力学仿真软件系统．利用软件系统进行堆芯物理计算，计算与验证结果表明，软件系统数学物理模型准确，可广泛应用于船用核动力装置模拟器的设计与研制。     

低浓化医院中子照射器（IHNI-1）堆芯的物理方案设计

采用蒙特卡罗程序MCNP／4B模拟计算了功率为30kW的低浓化医院中子照射器的堆芯物理参数,设计了合理的堆芯布置方案、235U富集度、控制棒价值、后备反应性和停堆深度,得到固有安全性较高、寿期达10年且无需换料、采用低浓化UO2燃料的医院中子照射器的堆芯物理设计方案,为后续反应堆工程设计以及硼中子俘获治疗肿瘤用中子束的设计提供理论依据。     

次锕系元素在加速器驱动的次临界快堆中嬗变的研究

选取加速器驱动次临界快堆（ADSFR）,进行嬗变来自于PWR（U）乏燃料中次锕系元素的研究。在堆芯内、燃料为NpAmCm的氧化物,选取液态钠为冷却剂。利用下列程序对所选方案进行物理计算和分析：LAHET－模拟质子与靶核的相互作用;MCNP4A－模拟次临界包层内20MeV以下的中子与材料核的相互作用;ORIGEN2－利用MCNP4A的输出提供的一群等效截面对堆芯进行燃耗计算。计算分析的结果表明：考虑临界安全、功率密度和燃耗等因素,利用所选方案进行次锕系元素嬗变是可行的。     

WIMS/CITATION在IHNI-1堆芯物理计算中的应用

建立了利用WIMS+CITATION计算医院中子照射器Ⅰ型堆堆芯中子学参数的模型,计算了堆芯的功率分布、顶铍反应性价值、控制棒价值、温度系数、堆芯燃耗等中子学参数,计算结果与文献数据一致,表明文章所建立的计算模型可用于医院中子照射器I型堆堆芯的物理计算。     

核石墨(NUCLEAR GRAPHITE)

核石墨是 40年代初 ,应建造核裂变反应堆的需要而研究发展出来的石墨材料的一个分支 ,在生产堆、气冷堆和高温气冷堆中用作慢化、反射和结构材料。石墨也是核聚变反应堆面向等离子体材料的很有希望的候选材料。铀的易裂变同位素Ｕ 2 35吸收中子后发生核裂变 :Ｕ 2 35 ｎ 2Ｆ 2ｎ～ 3ｎ 2 0 0ＭｅＶ核工程的任务就是安全、有效地利用核裂变释放出来的中子和巨大能量。中子与原子核发生反应的几率称之为截面 ,Ｕ 2 35的热中子 (平均能量为 0 .0 2 5ｅＶ)裂变截面比裂变中子 (平均能量为 2ｅＶ)裂变截面高两个等级。为了有效地利用裂变释放…     

The measurement of neutron and neutron induced photon spectra in fusion reactor related assemblies

The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out.The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the basis of FENDL-1 data. The agreement is better than 10% except for the photon fluence from the iron benchmark experiment, which is underestimated in the calculations by 25%.     

医院中子照射器I型堆超热中子束流孔道的优化设计

采用蒙特卡罗程序(Monte Carlo neutron and photo transport code,MCNP)对医院中子照射器Ⅰ型堆(IHNI-1)超热中子束流孔道的慢化层、反射层进行了优化设计。首先对FLUENTAL、Al等材料组成的6种慢化体方案进行了分析比较,给出了孔道出口处超热中子通量密度较大的两种设计方案;基于此两种慢化体设计方案,在保持束流孔道外框尺寸不变情况下,对慢化体周围的反射层进行了分析比较,给出了反射层的推荐方案;基于慢化体和反射层优化方案,最后给出了超热中子束流孔道出口处束流参数的空间分布。     

Personal neutron dosimetry at a research reactor facility

Individual neutron monitoring presents several difficulties due to the differences in energy response of the dosemeters. In the present study, an individual dosemeter (TLD) calibration approach is attempted for the personnel of a research reactor facility. The neutron energy response function of the dosemeter was derived using the MCNP code. The results were verified by measurements to three different neutron spectra and were found to be in good agreement. Three different calibration curves were defined for thermal, intermediate and fast neutrons. At the different working positions around the reactor, neutron spectra were defined using the Monte Carlo technique and ambient dose rate measurements were performed. An estimation of the neutrons energy is provided by the ratio of the different TLD pellets of each dosemeter in combination with the information concerning the worker's position; then the dose equivalent is deduced according to the appropriate calibration curve.     

基于WIMS和MCNP耦合程序的医院中子照射器I型堆燃耗计算

建立了基于WIMS和MCNP的临界-燃耗耦合计算方法,并对此方法进行了验算.通过栅元和组件问题的分析计算以及西安脉冲堆燃耗实验对比,验证了此耦合程序的可靠性和正确性.最后应用此耦合程序对医院中子照射器Ⅰ型堆的燃耗进行了计算和分析.     

Hardening neutron spectrum for advanced actinide transmutation experiments in the ATR

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast test reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas release modelling, needs to be accurately predicted and the hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are performed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neutron spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region.     

TORT/MCNP coupling method for the calculation of neutron flux around a core of BWR

For the analysis of BWR neutronics performance, accurate data are required for neutron flux distribution over the In-Reactor Pressure Vessel equipments taking into account the detailed geometrical arrangement. The TORT code can calculate neutron flux around a core of BWR in a three-dimensional geometry model, but has difficulties in fine geometrical modelling and lacks huge computer resource. On the other hand, the MCNP code enables the calculation of the neutron flux with a detailed geometry model, but requires very long sampling time to give enough number of particles. Therefore, a TORT/MCNP coupling method has been developed to eliminate the two problems mentioned above in each code. In this method, the TORT code calculates angular flux distribution on the core surface and the MCNP code calculates neutron spectrum at the points of interest using the flux distribution. The coupling method will be used as the DOT-DOMINO-MORSE code system. This TORT/MCNP coupling method was applied to calculate the neutron flux at points where induced radioactivity data were measured for 54Mn and 60Co and the radioactivity calculations based on the neutron flux obtained from the above method were compared with the measured data.     

Photon and neutron dose discrimination using low pressure proportional counters with graphite and A150 walls

A graphite-walled proportional counter with low neutron sensitivity was used in combination with a tissue-equivalent proportional counter (TEPC) to separate the photon and neutron components in mixed radiation fields. Monte Carlo (MCNP4C) simulations of the photon and neutron responses of the two detectors were done to obtain correction factors for the sensitivity differences. In an alternative method the radiation components were determined using constant-y(D)-values for typical photon and neutron energy distributions. The results show no significant difference between the two methods and the measured neutron dose-equivalent agrees within +/-50% with Bonner sphere determined values. The experimental data were obtained in measurement campaigns organised within the EVIDOS-project.     

Characterisation of the IRSN graphite moderated Americium-Beryllium neutron field

The SIGMA facility was set up at IRSN to provide thermal neutrons for metrology and dosimetry purposes. SIGMA consists of six Am-Be radioactive sources located in a 1.5 x 1.5 x 1.5 m3 graphite moderator block. The neutron field at the calibration position, situated at 50 cm from the west surface of the assembly was characterised experimentally and by Monte Carlo calculations. The thermal neutron fluence was determined by the activation of gold foils; the neutron fluence energy distribution above 240 keV was measured with proton recoil spectrometers and the neutron fluence energy distribution from thermal energies to 20 MeV was measured with a Bonner spheres spectrometer. A Monte Carlo simulation of the SIGMA assembly was undertaken using the MCNP4C code, and the calculated neutron fluence energy distribution was compared with the measurements. As a whole, the experimental data and the MCNP calculation are in a good agreement.     

Study on beam geometry and image reconstruction algorithm in fast neutron computerized tomography at NECTAR facility

Investigations on the fast neutron beam geometry for the NECTAR facility are presented. The results of MCNP simulations and experimental measurements of the beam distributions at NECTAR are compared. Boltzmann functions are used to describe the beam profile in the detection plane assuming the area source to be set up of large number of single neutron point sources. An iterative algebraic reconstruction algorithm is developed, realized and verified by both simulated and measured projection data. The feasibility for improved reconstruction in fast neutron computerized tomography at the NECTAR facility is demonstrated.