Evaluation of the fraction of delayed photoneutrons for TMSR-SF1

The 10 MW_(th) solid-fueled thorium molten salt reactor(TMSR-SF1) is a FLi Be salt-cooled pebble bed reactor to be deployed in 5–10 years, designed by the TMSR group. Due to a large amount of beryllium in the core, the photoneutrons are produced via(γ , n) reactions.Some of them are generated a long time after the fission event and therefore are considered as delayed neutrons. In this paper, we redefine the effective delayed neutrons into two fractions: the delayed fission neutron fraction and the delayed photoneutron fraction. With some reasonable assumptions, the inner product method and the k-ratio method are adopted for studying the effective delayed photoneutron fraction. In the k-ratio method, the Monte Carlo code MCNP6 is used to evaluate the effective photoneutron fraction as the ratio between the multiplication factors with and without contribution of the delayed neutrons and photoneutrons. In the inner product method, with the Monte Carlo and deterministic codes together, we use the adjoint neutron flux as a weighting function for the neutrons and photoneutrons generated in the core. Results of the two methods agree well with each other, but the k-ratio method requires much more computing time for the same precision.     

多传感器的数据融合落点定位方法研究

为了提高靶场末区落点定位的可靠性和准确性,提出了一种多传感器测量数据融合的定位方法。在分析了各传感器测量原理的基础上,以光电经纬仪和视听测量组成融合系统,给出这两类传感器数据融合定位的计算方法,并利用实测数据进行了计算,结果表明,该方法是切实可行的。     

Analysis on reactivity initiated transient from control rod failure events of a molten salt reactor

In a molten salt reactor （MSR）, the fuel is dissolved in fluoride salt. In this paper, the reactivity worth and reactivity initiated transient of Molten-Salt Reactor Experiment （MSRE） in the control rod failure events are analyzed, The point kinetic coupling heat-transfer model with decay character of six-group delayed neutron precursors due to the fuel motion is applied. The relative power and temperature transient under reactivity step and ramp initiated at different power levels are studied. The results show that the reactor power and temperature increase to a maximum, where they begin to decrease to stable values. Comparing with full power level, the transient result at low power level is more serious. The results are of help in our study on safety characteristics of an MSR system.     

Impact of cross section libraries on CASMO-4E and CASMO-4 results

A detailed comparison between the code CASMO-4 and its extended version CASMO-4E has been made. In addition to the standard library, CASMO-4E calculations have been performed also with its extended libraries. The differences are significant enough to be considered when choosing the library to be used for a particular problem. The differences in the multiplication factor k_∞ range up to several hundred pcm depending on the void history, burnup and other parameters. The differences in fuel temperature or void coefficients are smaller especially at small void fraction and low burnup. At large void and low burnup CASMO-4E with the standard library gives significantly different results than the other combinations. The microscopic cross sections show small differences when calculated with the same library but clear differences due to the extended libraries.     

Analyses of the V1000CT-1 benchmark with the DYN3D/ATHLET and DYN3D/RELAP coupled code systems including a coolant mixing model validated against CFD calculations

Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and three out of four MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical MCP characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. For an adequate modelling of the redistribution of the coolant flow in the reactor pressure vessel during the transient a simplified mixing model for the DYN3D/ATHLET code was developed and validated against a computational fluid dynamics calculation.

The results of both coupled code calculations are in good agreement with the available experimental data. The discrepancies between experimental data and the results of both coupled code calculations do not exceed the accuracy of the measurement data. This concerns the initial steady-state data as well as the time histories during the transient. In addition to the validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermal hydraulic models of the system codes RELAP5 and ATHLET and to explain differences between the calculation results.     

Preliminary safety analysis on loss of flow accidents and external source transients for LBE cooled ADSR core

In this paper, preliminary safety studies on the 800 MWth accelerator-driven system (ADS) proposed by Xi'an Jiaotong university are presented. The system is a pool type facility coupling a proton accelerator with current in the range of 17–23 mA and a sub-critical core by means of a spallation target. The RELAP5/MOD3.3 code is selected as a base tool. In order to simulate the system, the point kinetics model is modified and the property of lead-bismuth is implemented to meet the requirement of ADS analysis. This paper focuses on the assessment of its response to the loss of flow events. The first part is originated from the failure of the pump and the second part derives from the significant flow blockage at a fuel assembly inlet. The reactivity insertion accidents are caused by the change of the proton beam current. The results show that the safety and criteria are satisfied and the system is tolerant to the loss of flow accidents and proton beam doubled accident and is sensitive to the external neutron changing.     

Simple Method for Evaluation of [lbar] Used for Kinetic Calculations of One Point Reactor Model

Graphite Material has been widely used as first wall material in present large tokamak devices. However, overall evaluation with respect to vacuum engineering properties, interactions with plasmas, and thermal and mechanical properties has not been systematically performed so far, though these properties much depend on the kind of the graphite material. For the overall evaluation of the graphite as fusion first wall material, the “Graphite Project” was organized in 1986 under the support of the Ministry of Education, Science and Culture. More than 20 institutions participated in this project and 27 graphite material (isotropic graphite, pyrolytic carbon, C/C composite) supplied from 8 graphite manufactureres of Japan were characterized as “common samples”.

It was found that the vacuum engineering properties such as outgassing, effective surface area and hydrogen permeation significantly depended on the pore structure of the graphite. Both the outgassing quantity and the surface area were observed to be small for the graphite with low density. The mechanism of hydrogen permeability was explained by the molecular flow through the pore structure. The chemical sputtering yield of metal deposited graphite was significantly lower than that of the graphite with clean surface. The hydrogen retention was considerably reduced by the iron or titanium deposition onto the graphite surface. The yield of the radiation enhanced sublimation for the isotropic graphite was measured and the result showed that the yield of the isotropic graphite was quite similar to that of the pyrolytic carbon. The heat load experiments showed that most of the isotropic graphite failed at roughly the same heat load and the fracture toughness was also approximately the same. The C/C composite materials, however, had the thermal shock resistance and the fracture toughness, which were several times larger than that of the isotropic graphite.     

凸度测量过程中点扩展函数的蒙特卡罗模拟与解析表达研究

散射校正是高精度凸度测量中的技术难点。本文采用点扩展函数(PSF)理想模型估计的方法,建立了点扩展函数的蒙特卡罗(MC)计算模型,分析了PSF随射线能量和钢板厚度的变化关系,并提出用多个单能的线性组合来代替连续能谱计算的方法,取得了较好的效果。采用双高斯函数模型,结合蒙特卡罗模拟的计算结果,对凸度测量系统的单能点扩展函数进行了解析求解,并给出0.18MeV射线下,凸度测量系统PSF的解析公式,为凸度测量系统的散射校正奠定了基础。     

Fast data acquisition system based on digital oscilloscopes for fluctuation measurements in a long pulse JT-60U tokamak plasma

We have developed a new data acquisition (DAQ) system with fast sampling rate for fluctuation measurements in a long pulse JT-60U tokamak plasma. This system is based on a powerful digital oscilloscope, which has a large acquisition memory up to 50 Mwords/ch, 1 MHz sampling rate and 16 bits high resolution AD convertors. The system is composed of plural digital oscilloscopes and mass storages. On this system, most of data acquisition processes are executed at each digital oscilloscope. This feature of the system leads to an advantage that the processing loads are distributed among the digital oscilloscopes. This system has been successfully employed for measurements of various fluctuations obtained through magnetic probes, beam emission spectroscopy and so on. The size of the acquired data using this system has reached up to 10 GB/shot so far. It has demonstrated that this system is very powerful for data acquisition of multi-channeled signals with high time resolution in a long pulse plasma.     

调节闪烁室法测量222Rn、220Rn时仪器工作状态对探测效率的影响

针对闪烁室法^222Rn、^220Rn的测量装置，介绍了一种调节仪器最佳工作状态的比较完善的方法，并用实验进行了探测效率比较。结果表明，通过调节闪烁室法测量此^222Rn、^220Rn浓度装置的工作状态，可以提高仪器的探测效率约10％。     

Impact of cross-section generation procedures on the simulation of the VVER-1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3D thermal-hydraulics/neutron kinetics models

In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Enerige Atomique (CEA), France a coupled three-dimensional (3D) thermal-hydraulics/neutron kinetics benchmark was defined. The overall objective of OECD/NEA V1000CT benchmark is to assess computer codes used in analysis of VVER-1000 reactivity transients where mixing phenomena (mass flow and temperature) in the reactor pressure vessel are complex. Original data from the Kozloduy-6 Nuclear Power Plant are available for the validation of computer codes: one experiment of pump start-up (V1000CT-1) and one experiment of steam generator isolation (V1000CT-2). Additional scenarios are defined for code-to-code comparison. As a 3D core model is necessary for a best-estimate computation of all the scenarios of the V1000CT benchmark, all participants were asked to develop their own core coupled 3D thermal-hydraulics/neutron kinetics models using the data available in the benchmark specifications and a common cross-section library. The first code-to-code comparisons based on the V1000CT-1 Exercise 2 specifications exhibited unacceptable discrepancies between two sets of results. The present paper focuses on the analysis of the observed discrepancies. The VVER-1000 3D neutron kinetics models are based on cross-section data homogenized on the assembly level. The cross-section library, provided as part of the benchmark specifications, thus consists in a set of parameterized two group cross sections representing the different assemblies and the reflectors. The origin of the observed large discrepancies was found mainly to lie in the methods used to solve the diffusion equation. The VVER reflector properties were also found to enhance discrepancies by increasing flux gradients at the core/reflector interface thus highlighting more the difficulties in some codes to handle high exponential flux gradients. This paper summarizes the different steps applied to analyze the neutronic codes and their predictions as well as the impact of cross-section generation procedures.     

300~#-1靶片内热中子注量率分布测量实验研究

介绍了以4 mm× 0 .0 0 5mm的金箔为探测片 ,采用绝对与相对测量法测量 30 0 #- 1靶片内部热中子注量率分布的原理和方法。给出了靶片所在堆芯位置的绝对中子注量率和靶片内相对中子注量率分布。为靶片自屏因子的确定提供了测量数据。     

Impact of edge localized mode on photo-elastic-modulator based motional Stark effect polarimetry

Intermittent increase in background light induced by edge localized mode (ELM) in high confinement mode (H-mode) plasma gives disturbances to polarization angle of incident light measured by motional Stark effect (MSE) polarimeter using photo-elastic modulator (PEM), even if the background light is not polarized. The error in the polarization angle γ_PEM determined by the light intensity at the frequency modulated by the PEM originates in a broadband frequency spectrum of the intermittent background light and applies to PEM based polarimeter in general. The impact of the background light induced by the ELM on the error in γ_PEM is numerically investigated in terms of the polarization angle of incident light γ, the peak intensity of background light normalized by the intensity of beam emission X_ELM, ELM frequency f_ELM, ELM decay time τ_ELM and the time constant of low-pass filter (LPF) per single stage τ_LPF. The error increases almost proportionally with γ between ±22.5° when inphase spectral amplitude is used in γ_PEM calculation. The error becomes quite large for γ approaching to the direction of PEM axis (±22.5°), when absolute spectral amplitude is used. The error increases with X_ELM but does not depend on τ_ELM. When the ELM frequency is sufficiently lower than a critical ELM frequency that is given by inverse of an effective LPF time constant (τ_LPF times the number of LPF stages), the error has its peak, the height of which is independent of f_ELM, soon after the ELM and vanishes between subsequent ELMs. However, when f_ELM is higher than the critical ELM frequency, errors induced by subsequent ELMs pile up and the error increases with f_ELM. In case of no pileup, the maximum error becomes larger for smaller τ_LPF, because impact of ELM temporally concentrates in a shorter period just after the ELM. The error induced by ELMs is roughly estimated to be ∼0.002° in a type-I ELMy H-mode JT-60U discharge, which is much smaller than other errors coming from calibrations (about 0.1-0.2°) and temporal fluctuation (less than about 0.05°), mainly since the background light induced by ELMs is weak compared to the beam emission X_ELM ∼0.07 in JT-60U. However, MSE diagnostics experiencing large background light (X_ELM) should take this kind of error into account.     

核动力反应堆零功率物理试验的多普勒发热点测量试验

为了定量确定零功率物理试验功率的上限,在反应堆零功率物理试验中,利用数字反应性仪测量多普勒发热点,以确定试验功率的范围、保证试验精度。叙述了本次多普勒发热点测量试验的原理、试验仪器、试验方法、试验结果及数据处理方法等,试验结果表明:利用数字反应性仪测得的反应性经过修正后可以准确地判断多普勒发热点,可为后续物理试验提供参考。     

堆芯流量变化对压力容器水位测量影响分析

CPR1000核电厂在每次换料大修期间需执行CCMS(Core Cooling and Monitoring System)校验试验,以获得计算压力容器水位L_(VSL)所需的堆芯动态压头损失系数,完成该试验耗时较长。论文依据调试和换料大修期间一回路冷却剂流量的变化情况评估堆芯动态压头损失系数的变化,并定量评价对L_(VSL)测量的影响。分析结果表明,在回路水力特性未发生明显变化的情形下,对L_(VSL)测量引入的误差很小。建议在L_(VSL)测量不确定度评定时引入堆芯动态压头损失变化的影响,在换料大修时校验流量变化对堆芯动态压头损失的影响是否在允许范围之内,可简化CCMS校验试验,提升机组的经济性。     

Nb-1Zr与1Cr18Ni9Ti高温真空钎焊工艺研究

采用BNi-5钎料,对Nb-1Zr与1Cr18Ni9Ti高温真空钎焊工艺进行试验研究,对试验钎焊的接头进行了密封性能检验、抗热冲击性能试验、接头界面区微观分析以及钎缝拉伸强度试验.通过分析试验结果,确定了Nb-1Zr与1Cr18Ni9Ti高温真空钎焊工艺规范参数.     

1AF料液体积精确计量系统研究

在带有吹气液位和密度测量系统的模拟计量槽上进行了若干批体积标定。将原始数据标准化到某参考条件(主要是温度)下，根据计量槽内横截面积的变化及其反映出的残差分布图和增量斜率图，将数据划分成若干个模型段。将同一段内的各批数据进行适当的修正和补偿，把体积作为自变量、液位作为因变量，采用反向回归模式进行处理，根据数据的分布情况求出合适的标定方程，然后求标定方程的反方程，得出体积测量方程，并计算出相应的不确定度。利用测量方程可测定生产中与实测液位相应的料液体积。该系统经过对比验证，体积测量的不确定度低于0．3％(k=3)。     

软X射线功率谱仪测量系统在"强光一号"的丝阵实验

闪烁体和光电管组成的软X射线功率谱测量系统,在西北核技术研究所的"强光一号"加速器上,测量得到了各种丝阵靶的X射线辐射功率波形,并对测量结果进行了分析.     

海阳核电1号机组倒送电辅助变压器冲击试验保护分析

厂用母线受电是核电工程建设中的重要里程碑节点,也是海阳核电一期工程联合队首次开展的大型综合调试启动项目,该里程碑节点的成功实现,标志着厂内能够获得稳定的电源,为后续即将展开的大面积调试提供了可靠的电源保障.文章介绍了海阳核电1号机组厂用母线受电期间辅助变压器冲击过程,对冲击过程中的临时保护定值计算和母联电流波形进行初步分析.