Shut-down dose rate analyses for the ITER electron cyclotron-heating upper launcher

The electron cyclotron resonance heating upper launcher (ECHUL) is going to be installed in the upper port of the ITER tokamak thermonuclear fusion reactor for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). The paper reports the latest neutronic modeling and analyses which have been performed for the ITER reference front steering launcher design. It focuses on the port accessibility after reactor shut-down for which dose rate (SDDR) distributions on a fine regular mesh grid were calculated. The results are compared to those obtained for the ITER Dummy Upper Port. The calculations showed that the heterogeneous ECHUL design gives rise to enhanced radiation streaming as compared to the homogenous dummy upper port. Therefore the used launcher geometry was upgraded to a more recent development stage. The inter-comparison shows a significant improvement of the launchers shielding properties but also the necessity to further upgrade the shielding performance. Furthermore, the analysis for the homogenous dummy upper port, which represents optimal shielding inside the launcher, demonstrates that the shielding upgrade also needs to include the launcher's environment.     

Fatigue assessment of the ITER TF coil case based on JJ1 fatigue tests

The material of the TF coil case in the ITER requires to withstand cyclic electromagnetic forces applied up to 3 × 10⁴ cycles at 4.2 K. A cryogenic stainless steel, JJ1, is used in high stress region of TF coil case. The fatigue characteristics (S–N curve) of JJ1 base metal and welded joint at 4.2 K has been measured. The fatigue strength of base metal and welded joint at 3 × 10⁴ cycles are measured as 1032 and 848 MPa, respectively. The design S–N curve is derived from the measured data taking account of the safety factor of 20 for cycle-to-failure and 2 for fatigue strength, and it indicates that an equivalent alternating stress of the case should be kept less than 516 MPa for the base metal and 424 MPa for the welded joint at 3 × 10⁴ cycles. It is demonstrated that the TF coil case has enough margins for the cyclic operation. It is also shown the welded joint should be located in low cyclic stress region because a residual stress affects the fatigue life.     

基于蒙特卡罗方法的航空机组人员辐射剂量评估北大核心CSCD

采用辐射仿真人体面元模型和数字化飞机模型,基于蒙特卡罗方法开展了航空机组人员的航空辐射剂量研究。选取昆明至北京航线为例,评估了航空机组人员在该航线飞行时受到的航空辐射有效剂量率,分析了人体参数对航空辐射有效剂量率的影响,并探索了航线参数变化时机组人员受到的航空辐射有效剂量率的变化。结果表明,机组人员在昆明至北京航线受到的航空辐射有效剂量率为2.114μSv/h,基于中国人参考生理特征的体模和高加索人体参数体模的有效剂量率评估结果差异为25.3%;航线参数中航线的高度是最主要的影响因素,14 km飞行高度的航空辐射有效剂量率达到10 km高度时的1.8倍,同时,航线纬度升高时,机组人员受到的航空辐射有效剂量率也会产生显著的提升。该研究对航空机组人员的辐射剂量评估具有一定的参考和指导意义。     

NODHEX: A 2-D diffusion code based on the nodal expansion method in hexagonal geometry and its validation

A 2-D neutron diffusion theory computer code NODHEX for hexagonal geometry has been developed. The nodal algorithm is based on the nodal expansion method proposed by Lawrence. The nodal equation formulation is accomplished by using a second-order polynomial approximation for the flux. The equations include additional terms of discontinuity which occur in the expression of transverse leakage for the hexagonal geometry, unlike the nodal equations (using a second-order polynomial approximation) formulated by Lawrence. The code has been validated by comparing its predictions for the SNR-300 and VVER-1000 benchmarks with the results of other standard computer codes like DIF3D and SNAP. The inclusion of the additional terms of discontinuity is found to improve the predictions relative to Lawrence's predictions, though the same second-order polynomial approximation was used for solving the nodal equations.     

Benchmarking the MCR2S system for high-resolution activation dose analysis in ITER

The MCR2S system is a tool for the analysis of activation dose in ITER and fusion devices based on the rigorous-2-step formalism, MCNP mesh-tallies and the FISPACT nuclear inventory code. It couples transport and activation calculations in an integrated manner, allowing for the necessarily fine energy and spatial resolutions required for this kind of analysis and overcoming limitations of previous similar tools. The system is routinely used at UKAEA to assist the design of several ITER components such as the ICRH heating system and LIDAR diagnostic. Here we describe its functionality and a first benchmark exercise performed using the fusion-relevant Frascati Neutron Generator shutdown dose rate experiment results. Neutron spectra, reaction rates, photon spectra and activation decay dose rates at several positions and times, obtained using FENDL-2.1/MC and EAF-2007 data, are presented and compared to those of earlier software and libraries, showing good agreement. C/E values close to unity are obtained in all cases, illustrating the robustness and reliability of this novel software.     

SP3-coupled global variance reduction method based on RMC code

The beam windows of high-energy beam lines are important,and it is sometimes difficult to design because it is necessary to ensure particle propagation with minimum disturbance and fulfill mechanical requirements at the same time.The upstream decay pipe window of the long baseline neutrino facility at Fermilab has an extremely large diameter(1.8 m),with a thickness of only 1.5 mm to separate the helium atmosphere in the decay pipe and the nitrogen atmosphere on the other side.Furthermore,the center of this dish-shaped window is expected to be a 200-mm-diameter beryllium dish welded to the outside aluminum alloy A6061,and this welded combination must withstand extreme conditions of a 2.4-MW,high-energy proton beam without leakage.These severe conditions make the design of this window an unprecedented chal-lenge.This paper describes the static thermal-structural analyses based on which the structure has been optimized,as well as dynamic analyses for understanding the shock-wave effects originating in the beam.After optimization,the maximum von Mises stresses in the window decreased significantly in both normal operation and accident cases,making our design very reasonable.     

SP3-coupled global variance reduction method based on RMC code

A global variance reduction(GVR)method based on the SPN method is proposed.First,the global multi-group cross-sections are obtained by Monte Carlo(MC)global homogenization.Then,the SP3 equation is solved to obtain the global flux distribution.Finally,the global weight windows are approximated by the global flux distribution,and the GVR simulation is performed.This GVR method is implemented as an automatic process in the RMC code.The SP3-coupled GVR method was tested on a modified version of C5G7 benchmark with a thick-ened water shield.The results show that the SP3-coupled GVR method can improve the efficiency of MC criticality calculation.     

The FEM based calculation of crack-tip strain rate for determining the crack growth rate of 304 stainless steel in BWR environments

This research calculates the crack growth rate (CGR) of the cooling pipe in the nuclear power plant. The crack is caused by the tension left in welding as well as by the coolant chemicals. There is, by far, no satisfactory deterministic model that can predict the crack growth rate. Although Macdonald's CEFM model is more close to a deterministic one, it requires accurate calculation of the crack-tip strain rate before the crack growth rate can be obtained by using Faraday's equation. The empirical equation used by the CEFM model for estimating the crack-tip strain rate is a drawback in making it a totally deterministic model. This research is set to make up this critical defect by applying non-linear finite element method to calculate the crack-tip strain rate. According to the metallic property, there is a plastic zone forming around the crack-tip. This research found when applying finite element method, the number of the elements (namely the mesh density) in the vicinity of the crack-tip has significant effects on the crack-tip strain obtained. Therefore, this research pioneered by using different mesh density in the vicinity of the crack-tip for the calculation of the crack-tip strain, the convergence of which is used for selecting the appropriate mesh density. The selected mesh density is then used to determine the dimension of the plastic zone, in which the crack-tip strain rate can be calculated. When the FEM results are compared with the experimental data, the nonlinear finite element method combined with CEFM shows a very satisfactory prediction capability.     

我国南方某核厂址蟾蜍剂量率评估的比较研究

选择我国南方某核厂址两栖动物——蟾蜍作为参考生物,建立了蟾蜍的生物解剖学模型和外照射环境模型。采用蒙特卡罗模拟技术计算源介质中放射性核素¹³⁷Cs、⁹⁰Sr和²³⁹Pu对靶组织/器官的辐射剂量率,由此计算蟾蜍整体的辐射剂量率。采用ERICA程序和RESRAD-BIOTA程序计算蟾蜍的剂量率,并与解剖学模型进行比较。结果表明:三种方法计算的蟾蜍内照射剂量率基本一致;由于外照射环境模型的不同,外照射剂量率估算结果并不相同,ERICA程序与解剖学模型计算的外照射剂量率结果更接近;解剖学模型关注生物组织/器官的辐射剂量,对于核素分布不均匀的生物个体研究具有重要的意义。     

The structure analysis of ITER cryostat based on the finite element method

In the ITER project the cryostat is one of the most important components. Cryostat shall transfer all the loads that derive from the TOKAMAK inner basic machine, and from the cryostat itself, to the floor of the TOKAMAK pit (during the normal and off-normal operational regimes, and at specified accidental conditions). This paper researches the dynamic structure strength of the ITER cryostat during the operation of TOKAMAK. Firstly the paper introduces the types of loads and the importance of every type load to the research. Then it gives out the method of building model and principle of simplified model, boundary conditions and the way of applying loads on the cryostat. Finally the author discussed the analysis result and the strength questions of cryostat, also, the author pointed out the opinions according to the analysis results.     

Investigating the burning characteristics of electric cables used in the nuclear power plant by way of 3-D transient FDS code

Burning characteristics of electrical cables are one of the key parameters for the fire hazard assessment of nuclear power plants (NPPs) since the cables are the essential sources of fire in the plants. A three-dimensional (3-D) transient computational fluid dynamics (CFD) code_FDS is adopted in this paper to simulate these characteristics related to the cable burning. Being one of the NRC licensing fire codes, the FDS includes the thermal-hydraulic equations, the turbulence model and the chemical combustion model, etc. In order to assess the CFD fire models used in this code, a burning test using the control cable with the outer jacket of polyvinylchloride (PVC) and the inner insulation of cross-linked polyethylene (XLPE) is conducted. The measured parameters associated with the burning characteristics include the heat release rate (HRR), O₂ depletion, and CO and CO₂ production, etc. Except the amount of O₂ consumption, the predicted transient behaviors of other parameters can reproduce the measured data. Based on the chemical combustion model in the FDS code, this discrepancy may be essentially resulted from the default value of hydrogen fraction (H_frac) contained in the soot since the soot yield for the burning of PVC material is high enough that the uncertainty in the H_frac value has a prominent effect on the amount of O₂ consumption. This explanation can be confirmed by a benchmark calculation for simulating a burning test with the polymethylmethacrylate (PMMA) fuel of low-soot yield. The present simulation works can provide the useful information for the plant staff or the researcher as they would perform the fire hazard analysis in the NPPs using the FDS code.     

Modeling AC losses in the ITER NbTi poloidal field full size joint sample (PF-FSJS) using the THELMA code

A first validation of the full version of the thermal–hydraulic electromagnetic (THELMA) code, developed for the analysis of transients in the cable-in-conduit conductors and coils relevant for the International Thermonuclear Experimental Reactor (ITER) is presented here. THELMA includes electromagnetic models of the cable joints and terminations (lumped parameter) and of the conductor (distributed parameter), while for the thermal–hydraulics of the helium coolant it includes a compressible 1D flow model. The AC losses induced by a pulsed coil in the NbTi poloidal field full size joint sample (PF-FSJS) right leg conductor, tested in 2002 at the Sultan facility in Villigen (CH), are considered as test bed for this exercise. The computed energy deposition and evolution of the temperature downstream of the heated zone are in good agreement with the measured values. However, the inter-bundle electrical conductances needed in input by the code are compatible with measured values only when a sufficiently refined model is used in the cable cross-section.     

Harmonics analysis of the ITER poloidal field converter based on a piecewise method

Poloidal field (PF) converters provide controlled DC voltage and current to PF coils.The many harmonics generated by the PF converter flow into the power grid and seriously affect power systems and electric equipment.Due to the complexity of the system,the traditional integral operation in Fourier analysis is complicated and inaccurate.This paper presents a piecewise method to calculate the harmonics of the ITER PF converter.The relationship between the grid input current and the DC output current of the ITER PF converter is deduced.The grid current is decomposed into the sum of some simple functions.By calculating simple function harmonics based on the piecewise method,the harmonics of the PF converter under different operation modes are obtained.In order to examine the validity of the method,a simulation model is established based on Matlab/ Simulink and a relevant experiment is implemented in the ITER PF integration test platform.Comparative results are given.The calculated results are found to be consistent with simulation and experiment.The piecewise method is proved correct and valid for calculating the system harmonics.     

Shutdown dose rate assessment with the Advanced D1S method: Development,applications and validation

The present paper addresses the recent developments and applications of Advanced-D1S to the calculations of shutdown dose rate in tokamak devices. Results of benchmarking with measurements and Rigorous 2-Step (R2S) calculations are summarized and discussed as well as limitations and further developments. The outcomes confirm the essential role of the Advanced-D1S methodology and the evidence for its complementary use with the R2Smesh approach for the reliable assessment of shutdown dose rates and related statistical uncertainties in present and future fusion devices.     

剂量率切换辐照加速方法适用性的数值评估

为进行剂量率切换辐照加速方法适用性的数值评估,基于定量损伤物理模型和剂量率切换加速方法实验测试条件,以氢浓度、初始缺陷浓度和温度为变量,对金属绝缘层半导体(Metal insulator semiconductor,MIS)结构进行了系统的数值模拟。结果表明,剂量率切换辐照加速方法的适用性对双极器件氧化层内的氢浓度及缺陷浓度有很强的依赖。因此,剂量率切换加速方法的实验可能对部分器件和电路不再适用。     

Biasing transition rate method based on direct MC simulation for probabilistic safety assessment

Direct Monte Carlo (MC) simulation is a powerful probabilistic safety assessment method for accounting dynamics of the system.But it is not efficient at simulating rare events.A biasing transition rate method based on direct MC simulation is proposed to solve the problem in this paper.This method biases transition rates of the components by adding virtual components to them in series to increase the occurrence probability of the rare event,hence the decrease in the variance of MC estimator.Several cases are used to benchmark this method.The results show that the method is effective at modeling system failure and is more efficient at collecting evidence of rare events than the direct MC simulation.The performance is greatly improved by the biasing transition rate method.     

Stratified flow-induced air-ingress accident assessment of the GAMMA code in HTGRs

An air-ingress accident is a major safety issue pertaining to high-temperature gas-cooled reactors. To see the effect of a stratified flow, which is a multi-dimensional phenomenon that occurs in large broken pipes, we perform 1-D and 2-D air-ingress simulations in the guillotine break of the main coaxial pipe of a 600 MWth GT-MHR with the GAs multicomponent mixture transient analysis (GAMMA) code. We used a 2-D fluid volume to build the coaxial inlet pipe, the lower plenum of the reactor core, and the cavity and simplified the other components as 1-D fluid blocks. After the guillotine break of the main coaxial pipe, the air in the reactor cavity flows into the reactor core in four phases: the blow-down phase, the stratified flow phase, the molecular diffusion phase, and the natural convection phase. In the early stage of a broken pipe, the lower plenum region of the reactor is filled with air within 30 s by a density-driven airflow. In a 1-D simulation, the process of filling the lower plenum with air ingressed from a cavity caused by the diffusion process takes 30 min. However, after 30 s, the flow velocity of air ingressed into the broken pipe decreases and the diffusion phase eventually begins. The natural circulation in this scenario starts after more than 360 h for the 1-D simulation but fails to commence after more than 500 h for the 2-D simulation. The belated natural circulation in the 2-D simulation is mainly attributed to the slower diffusion process in the core region. In turn, the slower diffusion occurs because the temperature of the air in the lower plenum is lower in the 2-D simulation than in the 1-D simulation. The maximum core temperature in the 2-D simulation was by 60 °C lower than that in the 1-D simulation.     

基于CODAC的ITER电流引线测控系统

中国科学院等离子体物理研究所(Institute of Plasma Physical,Chinese Academy of Sciences,ASIPP)负责国际热核聚变实验堆(International Thermonuclear Experimental,ITER)60根高温超导电流引线(High Temperature Superconducting Current Lead,HTSCL)产品的研制与测试,并在2013–2015年间开展了三对三种电流等级(68 k A、55 k A和10 k A)的高温超导电流引线认证制造。为检验电流引线的低温大电流性能,ASIPP与印度塔塔咨询服务公司(Tata Consultancy Service,TCS)及ITER合作开发了基于CODAC(Control,Data Access and Communication)框架的ITER高温超导电流引线测控系统。该系统包括传统西门子PLC300工艺过程测控系统、基于Lab VIEW的失超保护系统、基于PLC400冗余设计的互锁系统和基于NI c系列模块的快速控制系统(Plant system Controller,Fast Controls,PCF)。目前本系统已通过三轮验收测试并在2015年1月份的ITER CC 10 k A电流引线原型件和同年7月份的ITER TF 68 k A电流引线原型件中成功应用。结果表明,本系统能很好地满足电流引线的实验需求,得到ITER国际认同。电流引线测控系统软硬件遵照ITER的CODAC标准进行设计,是CODAC和互锁保护规范的首次在ITER真实组件物理性能测试的联合应用案例,可作为ITER采购包出厂验收推行的CODAC模范。     

Development of a subchannel code for blockage accidents of LMFRs based on the 3D fuel rod model

To predict the thermal-hydraulic(T/H) parameters of the reactor core for liquid-metal-cooled fast reactors(LMFRs), especially under flow blockage accidents, we developed a subchannel code called KMC-FB.This code uses a time-dependent, four-equation, singlephase flow model together with a 3D heat conduction model for the fuel rods, which is solved by numerical methods based on the finite difference method with a staggered mesh. Owing to the local effect of the blockage on the flow field, low axia...