Study on kinetics of fast subcritical assembly—Contribution of delayed neutrons to the transition after a reactivity insertion in a subcritical system—

This study is an investigation of the effect of the delay neutron on the kinetics in the subcritical system. And, it proposes a method necessary for the kinetics code development that uses the Monte Carlo (MC) computation.

It is generally difficult to analyze three dimensional space and time dependent kinetics by using a MC method. It is because the sampling of the neutron in a region becomes difficult when conditions of the region changes with time. In this study, we consider about the effect of delayed neutron in the kinetics of ADS. The behavior of neutrons is considered spontaneous in this system. It means a neutron is absorbed or leaks in a short period, while the conditions of region do not change. Therefore they are treated by steady state calculation. On the other hand the densities of delayed neutron precursors changes slowly, and the conditions of region change. In the concept of developed MC method, the neutrons are calculated by using steady state equation at each time point, and the delayed neutron precursors are calculated by using time dependent equation. We tried to inspect the accuracy of this method by using a point equation. We obtained strict solution Φ* as a reference solution, Φ1 as a solution by the present method, and Φ2 as the solution where both neutrons and delayed neutron precursors are treated by using static equations. The obtained results show a good agreement between Φ1 and Φ*, though the Φ2 agrees with Φ* poorly in all cases. Especially, we showed that this technique was effective from the reactivity change by ADS, and the relation of a delayed neutron. Finally, the effect of the delay neutron on the beam trip in the neutron source for the drive was examined by using the technique of Φ2.     

基于IQS/MC方法的ADS次临界反应堆中子时空动力学模拟分析

采用改进准静态近似与蒙特卡罗中子输运程序相结合(IQS/MC)的方法实现了加速器驱动的次临界系统(ADS)中子时空动力学模拟计算。以加速器驱动嬗变研究装置的靶堆耦合参考方案物理模型为例,通过对束流瞬变引入和燃料组件提升两种工况进行动态模拟,计算得到了堆芯总的相对功率、分能群相对中子注量率及相对功率三维网格分布随时间的变化。将IQS/MC方法计算结果与点堆计算结果进行了对比分析,模拟结果符合物理规律,两种方法对比结果与国外相关文献一致,表明IQS/MC方法适用于ADS次临界反应堆中子时空动力学过程的瞬态安全分析。     

Measurement of kinetic parameters in the fast subcritical core MASURCA

In the MUSE shared cost action of the European Fifth Framework Program measurements have been performed to investigate the neutronic behavior of the fast subcritical core MASURCA coupled with the GENEPI accelerator. The aim is to examine the applicability of different measurement techniques for the determination of the main kinetic parameters. The measurement of Rossi-alpha distributions, recorded with the accelerator turned off, showed that the analysis of the obtained distributions is feasible for deep subcritical levels, but with strongly deteriorated statistics. From Rossi-alpha distributions, recorded with the pulsed neutron source in operation, the alpha decay constant was easily derived due to good statistics on the correlated signal resulting from the strong intensity of the neutron pulse. When applying the pulsed neutron source analysis, the reactivity (in dollars) together with the ratio of the mean neutron lifetime l and the effective delayed neutron fraction β_eff is immediately derived. Although these first results are very promising, further measurements are needed to qualify the method at larger subcritical levels which are representative for future ADS.     

Model independent determination of the elastic incoherent structure factor in neutron scattering experiments

In quasielastic neutron scattering (QENS) the dynamic structure factor S(Q, ω) consists of an elastic central line with area given by the elastic incoherent structure factor (EISF) and a continuous quasielastic broadening. Using only the resolution function of the spectrometer we discuss a method to derive the EISF directly from the raw experimental data points S(Q, ω) without any smoothing or any modelling of the quasielastic broadening. We compare this model to the Fourier transform method, show how it works on actual neutron scattering data and discuss its accuracy for several cases.     

Channeling of low energy light and heavy ions in single-wall nanotubes

The Monte Carlo simulation program has been used to study low energy channeling in the single-wall nanotube and its rope, in comparisons between beam sizes and between light (He) and heavy (Ar) ions. The simulation mainly shows that the critical angle Ψ_C = 48 E^−1/2 (E is incident energy) for the He (light) ion channeling but Ψ_C = 18 E^−1/2 for the Ar (heavy) ion channeling, in the (17,0) zigzag single-wall nanotube. Thus, it might be found in the simulation that Ψ_C strongly depends on the ion mass.     

Electron emission induced by H and H incident on thin carbon foils: influence of charge changing processes

The forward and backward electron emission yields γ_F and γ_B have been calculated by Monte Carlo simulation for protons (H⁺) and hydrogen atoms (H⁰) (with energies between 25 keV and 5 MeV) incident on thin amorphous carbon foils. Direct electron excitations by the incident projectiles as well as electron excitations resulting from charge exchange processes undergone by H⁺ or H⁰ have been taken into account. For the latter, Auger and Shell processes have been considered. Subsequent electron transport has been considered in order to calculate the forward and backward electron emission yields γ_F and γ_B.     

An efficient deterministic method for generating non-negative scattering cross-sections

In the conventional discrete ordinates approach, the scattering source is approximated as a truncated Legendre series. In case of highly anisotropic scattering problems (e.g., incident beam problems), the truncated Legendre scattering cross-sections give unphysical negative cross-sections in some values of μ (cosine of scattering angle). These unphysical artifacts cause negative scattering sources and negative angular fluxes. In addition to that, negative angular fluxes may cause wrong scalar flux as well. A new method to generate non-negative scattering cross-sections, which is very efficient and deterministic, is proposed in this paper. The main idea of this method is to make non-negative scattering cross-sections that produce equivalent scattering sources. This method does not have a practical limitation to generate non-negative scattering cross-sections because the calculations of the scattering sources are performed with the conventional truncated cross-section data provided from the standard processing codes. In the both neutron and the photon/electron coupled cases, an inadequate truncated Legendre order causes problems, e.g., inaccurate angular distribution of scattering for low order, oscillations of differential scattering cross-sections where cross-sections are small for high order expansions, and negative differential scattering cross-sections in some values of μ.

In the neutron case, if the anisotropy is very high compared to the truncated Legendre order, inaccurate angular distributions of scattering occur especially in within-group scattering. Even if the truncated Legendre order is quite high to represent angular distribution of scattering in the photon/electron coupled case, it still causes oscillations where the cross-sections are small.

The method in this paper achieves both an accurate angular distribution of scattering and non-negative scattering cross-sections for neutron and photon/electron coupled cases. The generated non-negative scattering cross-sections with the new method are compared to the conventional scattering cross-sections and tested in the transport calculations. The numerical results tested in the transport calculation give accurate results without unphysical negative angular fluxes.     

Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity (ρ_ex), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU).     

Isotope effects in thermal neutron transmission and backscattering processes for ε-phase zirconium hydrides and deuterides

Distributions of hydrogen isotope concentrations in ε-phase zirconium hydrides and deuterides (ε-ZrH_x and ε-ZrD_x: 1.8 < x < 2.0) were investigated by neutron radiography (NRG). The NRG images of the thermal neutron transmission and backscattering revealed hydrogen concentration dependence and isotope differences. The thermal neutron mass attenuation coefficients in relation to the hydrogen isotope concentrations were determined from the transmission NRG images. The results showed the isotope effects of the thermal neutron mass attenuation coefficients for ε-ZrH_x to be about 6–9 times higher than those for ε-ZrD_x. The neutron scattering processes for transmission and backscattering NRG images of ε-ZrH_x and ε-ZrD_x were also analyzed using a general Monte Carlo neutron-particle transport (MCNP) code.     

Influence of the target thickness on the backward and forward electron emission characteristics induced by protons incident on thin carbon foils

The forward (γ_F) and backward (γ_B) electron emission yields have been measured for protons incident on thin carbon foils for incident energies between 2 and 7 MeV as a function of the target thickness. Comparisons with theoretical results obtained by Monte Carlo simulations are presented. In particular, the Meckbach factor R_γ = γ_F/γ_B is discussed.     

Subcritical reactivity monitoring in accelerator driven systems

In this paper, an absolute measurements technique for the subcriticality determination is presented. The development of accelerator driven systems (ADS) requires the development of methods to monitor and control the subcriticality of this kind of system, without interfering with its normal operation mode. This method is based on the stochastic neutron and photon transport theory that can be implemented by presently available neutron transport codes. As a by-product of the methodology a monitoring measurement technique has been developed and verified using two coupled Monte Carlo programs. The first one, LAHET, simulates the spallation collisions and the high energy transport and the other, MCNPDSP, is used to estimate the counting statistics from neutron ray counter in fissile system, and the transport for neutrons with energies less than 20 Mev. Through the analysis of the counter detectors it is possible to determine the kinetics parameters and the k_eff value. We present two different ways to obtain these parameters using the accelerator or using a Cf-252 source. A good agreement between theory and simulations has been obtained with both sources.     

Diamagnetic measurement on HT-7 superconducting tokamak

Diamagnetic measurement is a basic diagnostics on tokamak. Some important plasma parameters such as plasma energy and betap β_p can be obtained from this measurement. For most case, diamagnetic flux ΔΦ is extremely smaller than toroidal flux Φ (ΔΦ/Φ − 10⁻⁴). Therefore we have to use techniques that allow measurement to better than 1 part in 104 to get 10% accuracy in value of β_p. Using a compensation coil is a typical technique to improve the signal to noise ratio. In this paper the design of diamagnetic diagnostics system for HT-7 superconducting tokamak device is introduced and some experimental results of plasma energy and β_p are given in different plasma discharges.     

Shielding design and analyses of KIPT neutron source facility

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of a neutron source facility. An electron accelerator drives a sub-critical facility (ADS) is used for generating the neutron source. The facility will be utilized for performing basic and applied nuclear researches, producing medical isotopes, and training young nuclear specialists. Monte Carlo code MCNPX has been utilized as the major design tool for the design, due to its capability to transport electrons, photons, and neutrons at high energies. However the ADS shielding calculations with MCNPX need enormous computational resources and the small neutron yield per electron makes sampling difficulty for the Monte Carlo calculations. The high energy electrons (E > 100 MeV) generate very high energy neutrons and these neutrons dominant the total radiation dose outside the shield. The radiation dose caused by high energy neutrons is ∼3-4 orders of magnitude higher than that of the photons. However, the high energy neutron fraction within the total generated neutrons is very small, which increases the sampling difficulty and the required computational time. To solve these difficulties, the user subroutines of MCNPX are utilized to generate a neutron source file, which record the generated neutrons from the photonuclear reactions caused by electrons. This neutron source file is utilized many times in the following MCNPX calculations for weight windows (importance function) generation and radiation dose calculations. In addition, the neutron source file can be sampled multiple times to improve the statistics of the calculated results. In this way the expensive electron transport calculations can be performed once with good statistics for the different ADS shielding problems. This paper presents the method of generating and utilizing the neutron source file by MCNPX for the ADS shielding calculation and similar accelerator facilities, and the accurate radiation dose analyses outside the shield using modest computational resources.     

Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.     

ADS次临界系统中子时空动力学计算与瞬态分析

加速器驱动次临界反应堆(ADS)中子时空动力学计算需要考虑外中子源和空间分布的影响,比临界系统中子动力学计算要复杂得多。本文将改进准静态(IQS)近似与蒙特卡罗(MC)方法相结合,对于带外源的ADS次临界系统中子时空动力学过程,形状函数、动力学参数由MCNPX程序计算得到,幅度函数与集总参数热工反馈模型进行耦合计算,并开发了IQS/MC计算程序可视化操作界面。针对CIADS靶堆耦合系统参考方案物理模型,对引入束流瞬变及无保护失流工况过程进行瞬态模拟计算分析,给出了堆芯相对功率、燃料温度及冷却剂出口温度随时间的变化曲线。同时,将中子注量率进行分群计算,得到了堆芯分能群的相对中子注量率网格分布随时间的变化,模拟结果与理论分析一致。     

Ion-induced electron emission – monitoring the structure transitions in graphite

The temperature dependence of ion-induced electron emission yield γ under 30 keV Ar⁺ ion impacts at incidence angles θ = 0−80° under dynamically steady-state conditions has been measured for polygranular graphite POCO-AXF-5Q. The fluencies were 10¹⁸–10¹⁹ ion/cm², the temperatures varied from the room temperature (RT) to 400 °C. The RHEED has shown that same diffraction patterns correspond to a high degree of disorder at RT. At high temperature (HT), some patterns have been found similar to those for the initial graphite surfaces. The dependence γ(T) has been found to be non-monotonic and for normal and near normal ion incidence manifests a step-like increase typical for a radiation induced phase transition. At oblique and grazing incidence (θ > 30°), a broad peak was found at T_p = 100 °C. An analysis based on the theory of kinetic ion-induced electron emission connects the behavior of γ(θ,T) to the dependence of both secondary electron path length λ and primary ion ionizing path length R_e on lattice structure that drastically changes due to damage annealing.     

MA含量对ADS堆芯动力学参数的影响研究

有效缓发中子份额（β_eff）、平均中子代时间（Λ）和反应性反馈系数（α）是核反应堆动力学中至关重要的参数。本文采用蒙特卡罗方法计算了加速器驱动的次临界系统（ADS）堆芯的动力学参数,并分析了次锕系核素（MA）装载量对这些参数的影响。通过在燃料中添加不同含量的MA,来研究其对ADS堆芯动力学参数的影响。结果表明,当MA在燃料中的含量从0%增加到5%时,β_eff和Λ的值分别降低了18%和31%,多普勒反馈系数平均值α^-D由-0.56 pcm/K变化到-0.36 pcm/K,冷却剂反馈系数平均值α^-C由-2.11 pcm/K变化到-1.73 pcm/K。     

Power density effect on feasibility of water cooled thorium breeder reactor

Breeding is made possible by the high value of neutron regeneration ratio η for ²³³U in thermal energy region. The reactor is fueled by ²³³U–Th oxide and it has used the light water as moderator. Some characteristics such as spectrum, η value, criticality, breeding performance and number density are evaluated. Several power densities are evaluated in order to analyze its effect to the breeding performance. The η value of fissile ²³³U obtains higher value than 2 which may satisfy the breeding capability especially for thermal reactor for all investigated MFR. The increasing enrichment and decreasing conversion ratio are more significant for MFR < 0.3. The required enrichment and conversion ratio do not change significantly caused by power density change for very tight lattice cell (MFR < 0.3), however, its strongly depends on the power density change for higher MFR (MFR ≥ 0.3). Breeding condition of all investigated power densities can be achieved for burnup ≥ 30 GW d/t at MFR = 0.3 and it requires about 3.5% of required ²³³U enrichment. Number density of ²³³Pa decreases significantly with decreasing power density which leads the reactor has better breeding performance because lower capture rate of ²³³Pa.     

蒙特卡罗方法在ADS屏蔽计算中的应用

利用蒙特卡罗方法计算了新一代核能系统加速器驱动系统(ADS)中质子束管内的中子归一化注量率分布以及通过质子束管入口和其它外表面逸出的归一化中子注量率,得出了一些对ADS系统的设计有重要意义的结论。     

含钆中子屏蔽材料的设计及其力学性能分析

设计了一种具有良好中子屏蔽能力、高强度及高韧性的新型中子屏蔽材料,用于吸收核电站乏燃料储存格架和乏燃料运输过程中的热中子辐射。材料通过蒙特卡罗粒子传输(输运）软件MCNP进行设计,并通过放电等离子烧结设备及热轧的方式制成了板材。MCNP模拟结果及材料热中子屏蔽测试结果表明：铝基Gd₂O₃复合材料的热中子屏蔽性能与铝基碳化硼相当。Gd₂O₃颗粒球磨后呈现μm、亚μm级甚至有些颗粒达到了nm级。随球磨时间的增加,材料的力学性能逐渐增强。X射线衍射检测发现了钆-铝合金相的生成。经TEM分析表明：材料的强化机制主要是位错强化和nm级Gd₂O₃颗粒的弥散强化,拉伸强度和伸长率分别达到了240 MPa和16%,其断口主要为韧性断裂。