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.     

Theoretical comparison of performance using transfer functions for reactivity meters based on inverse kinetic method and simple feedback method

The performance of two digital reactivity meters, one based on the conventional inverse kinetic method and the other one based on simple feedback theory, are compared analytically using their respective transfer functions. The latter one is proposed by one of the authors. It has been shown that the performance of the two reactivity meters become almost identical when proper system parameters are selected for each reactivity meter. A new correlation between the system parameters of the two reactivity meters is found. With this correlation, filter designers can easily determine the system parameters for the respective reactivity meters to obtain identical performance.     

基于空气冷却的熔盐堆非能动余热排出系统优化设计

上海应用物理研究所基于TRISO包覆球形颗粒燃料与液态氟盐提出了基于钍基熔盐固态试验堆（TMSR-SF1）技术方案,其中一个重要的工作是非能动余热排出系统（PRHRS）设计。由于熔盐与水的不兼容特性,以及其高运行温度,采用空气作为最终热阱来设计PRHRS成为必然。为实现系统最简化、体积最小化以及排热与保温兼顾的设计目标,本文从MSR堆芯活性区到外界空气热阱传热过程的模型入手,建立了PRHRS优化设计模型,获得了优化设计方案,并基于改进的RELAP5/MOD4.0程序（针对TMSR-SF1的专门改进程序）开展了PRHRS容量论证评价,经计算分析,PRHRS容量设计合理,可确保反应堆全厂断电（SBO）后排热安全。      

熔盐堆物理热工耦合程序开发及验证分析

熔盐堆(Molten Salt Reactor,MSR)是第四代反应堆6种堆型中唯一的液态燃料反应堆,与固态燃料-液体冷却剂反应堆相比,原理上有较大不同。在熔盐堆中,流动的熔盐既是燃料又是冷却剂与慢化剂,中子物理学与热工水力学相互耦合;由于熔盐的流动性,缓发中子先驱核会随燃料流至堆芯外衰变,造成缓发中子的丢失,导致堆芯反应性降低。正是由于熔盐堆的这些新特性,造成熔盐堆内缓发中子先驱核、温度等参数变化与固态燃料反应堆有所不同,需要研究熔盐堆在各种工况下的相关物理参数变化。本文主要工作是考虑缓发中子先驱核的流动性对熔盐堆的影响,研究适用于熔盐堆的二维圆柱几何时空中子动力学程序及与之耦合的热工水力学程序;利用该程序对熔盐堆中子物理学和热工水力学进行耦合计算,验证熔盐堆相关实验数据;并且计算了熔盐堆无保护启停泵及堆芯入口温度过冷过热工况,用于分析熔盐堆的安全特性。计算结果表明,程序能够对熔盐反应堆实验(Molten Salt Reactor Experiment,MSRE)的相关实验数据进行较好的模拟计算,并且验证了熔盐堆的固有安全性。     

Subcriticality measurement using time-domain decomposition-based integral method for simultaneous reactivity and source changes

ABSTRACT

To estimate the subcriticality in dollar units for an arbitrary state-change, the time-domain decomposition-based integral method (TDDI) is proposed using the point kinetics theory based on the fundamental mode approximation. In a general transient subcritical system, reactivity, neutron source intensity, and point kinetics parameters can vary simultaneously. Furthermore, the state-change may not necessarily be a stepwise change. For such a transient, the TDDI method can estimate the subcriticality after the transient using only the time variation of the neutron count rate. Therefore, the proposed method is useful to approximately estimate the subcriticality in a system where a detailed core configuration is unknown. To investigate the applicability of the TDDI method, transient experiments with simultaneous reactivity and source changes or to two successive safety rods dropping were performed at the Kindai University Training and Research Reactor (UTR-KINKI). By comparing with reference values using excess reactivity and control rod worth, it was validated that the subcriticality values obtained by the TDDI method better agree with the reference values than the previous integral method.     

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.     

Design & construction of a digital real time reactivity meter for Tehran research reactor

Reactivity measurement is one of the challenges of monitoring, control and investigation of nuclear reactors. In this paper design and construction of a reactivity meter for continuous monitoring of reactivity in research reactors are described. The device receives amplified output of the fission chamber, which is in mA range, as the input. Using amplifier circuits, this current is converted to voltage and then digitalized with a microcontroller to be sent to serial port of computer. The device itself consists of software, which is a MATLAB real time programming for the computation of reactivity by the solution of neutron kinetic equations. After data processing the reactivity is calculated and presented using LCD. Tehran research reactor is selected to test the reactivity meter device. The results of applying this reactivity meter in Tehran research reactor (TRR) are compared with the experimental data of control rod worth, void coefficient of reactivity and reactivity changes during approach to full power. The maximum relative error in several experiments is calculated to be 13%.     

随机球床模拟方法对氟盐冷却高温堆中子学参数的影响分析

氟盐冷却高温堆（FHR）采用氟盐冷却球形燃料元件,其中子物理计算面临双重不均匀性问题：燃料球在堆芯内的随机排布和包覆燃料颗粒在燃料球中的随机排布。此问题是该堆型设计中面临的主要挑战之一。本文基于MCNP程序和固态燃料钍基熔盐堆（TMSR-SF1）模型完成了不同燃料球床与燃料球描述对关键中子学参数（如k_eff、堆芯能谱、控制棒价值和温度系数等）的影响分析。燃料球床描述使用随机序列添加（RSA）方法建立了随机球床模型与体心立方（BCC）结构的等效规则模型。包覆燃料颗粒描述则基于简立方（SC）等效模型利用MCNP程序中的URAN卡实现随机扰动。结果表明,包覆燃料颗粒随机分布的影响远小于燃料球随机分布的影响;尽管具有相同的总堆积密度,等效规则模型相比于随机球床模型会增加堆芯中子的泄漏,低估冷态满装载反应性约0.5%,高估控制棒价值约5%。     

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.     

Novel approach to solving the inverse equation of point kinetics by the Bernoulli number generalisation method

ABSTRACT

A new approach to solve the inverse equation of point kinetics is presented. The neutron density history can be considered as a series of infinite Bernoulli numbers. In order to simplify the problem, the proposed method uses an approximation which considers only two derivatives of the neutron population density. The reactivity is calculated for different forms of neutron population density and for different time steps. Due to its high precision, the results obtained suggest that the proposed method can be used in a real-time digital reactivity meter.     

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.     

粒子物理实验中的精密时间间隔测量

本文是关于粒子物理实验中精密时间间隔测量的电子学方法和技术的一个综述.描述了精密时间间隔测量在粒子物理实验中的作用,以及粒子物理实验对时间间隔测量系统的性能指标需求.讨论了时间间隔测量的基本电子学手段,并着重分析了时间内插技术的分类和实现方法.     

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.     

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.     

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.