SARCS-4程序系统临界物理模拟实验验证

SARCS-4程序系统是中国核动力研究设计院自主研发的新一代中子学程序包,需对程序展开工程验证,完善理论模型,提高计算精度。利用成熟燃料元件,设计并制造出与新型燃料组件结构相似的模拟组件。利用模拟组件构造了3种堆芯布置并进行模拟实验,验证SARCS-4程序系统的正确性和可靠性。按照单变量准则和多样化准则,考察控制棒、可燃毒物棒、围板和堆芯布置等因素。模拟实验临界棒位校核分析表明:堆芯泄漏、围板效应、控制棒和可燃毒物棒效应是影响校核精度的主要因素,SARCS-4计算程序系统对模拟实验的整体计算精度相对较高,特殊布置堆芯仍需进一步提高计算精度,后续将通过进一步的实验和研究开展持续验证和改进。     

堆芯中子学程序系统SARCS-4.0的开发及初步验证

开发了堆芯中子学程序系统SARCS-4.0,该程序系统能处理由任意方形燃料组件组成的堆芯;能计算铀钚、钍铀燃料循环;能计算硼、钆、铒、铪、银、铟、铕、钐等各类可燃毒物和含硼、铪、银-铟-铬、铕、镝等各类控制棒;具备堆芯核设计的基本功能.使用SARCS-4.0系统对超临界水冷反应堆(SCWR)堆芯进行计算以验证程序系统的计算准确性,结果表明,SARCS-4.0系统具有较高计算精度,该系统从功能上、精度上均适用于新型反应堆堆芯选型研究.     

21 热离子反应堆电源中子物理模拟实验样机环境影响

热离子反应堆电源中子物理模拟实验样机（SPEM，简称模拟实验样机），是一座采用铀-235富集度90％的金属铀块作核燃料、氢化锆作慢化剂、金属铍作反射层、转动控制鼓控制反应性的临界装置。     

4 启明星1#次临界装置上第2阶段实验研究

启明星1#次临界装置建成后，在第1阶段的实验研究即用Am-Be稳态外中子源驱动启明星1#次临界装置，Am-Be稳态外中子源的平均中子能谱约4MV，初步测量了其中子学特性后，又于2005年10月到11月进行了第2阶段的实验，即用高压倍加器产生的脉冲外中子源驱动启明星1#次临界装置。     

2 ADS次临界中子学研究进展

ADS系统反应堆物理基础研究为中国科技部973项目“加速器驱动核能系统的物理及技术基础研究”的第2课题。五年来，已经完成了计划任务书中规定的所有的项目，包括第1阶段的稳态外源（^252Cf）驱动水堆零功率次临界（东风3号）实验及分析和第2阶段的建立“启明星”实验装置并开展初步实验研究。     

19 ERANOS程序用于CEFR堆芯物理特性的验证计算

ERANOS系统（欧洲反应堆分析优化系统）是欧洲和日本联合开发的快中子反应堆堆芯物理屏蔽计算软件系统，采用模块化设计，包含核截面库制作、中子学计算和燃耗计算等模块。该系统可进行反应堆中子学一维至三维的扩散、输运计算，可进行堆芯中子动态特性、燃料管理以及灵敏度分析等计算。     

17 CEFR外推临界程序离线验证

外推临界程序是CEFR物理启动软件中1个重要的子程序之一。在物理启动软件编制完成后，将分别对其中的各个子程序进行实验验证。本工作仅就外推临界程序的离线验证过程进行简单的介绍。     

临界实验装置水位外推达临界方法优化研究

带乏燃料的临界实验装置可用于开展乏燃料贮存相关的临界实验研究,考虑到其具有较高放射性,因此开展实验时最佳方法是水位外推达临界方法。目前该带乏燃料的临界实验装置尚未建成,遂使用与其堆芯结构、物理特性相似的铀棒栅轻水慢化临界实验装置开展水位外推达临界方法的模拟实验。在该实验中,由于水的屏蔽作用随水位上升不断加深导致无法得到正常的外推曲线。本文通过蒙特卡罗程序模拟外推过程,研究了中子源与探测器在不同的相对位置处中子计数率随水位变化的规律,给出了探测器与中子源的优化布置方案。此外,还提出一种以水位价值作为外推参数的外推方法,消除了水位高度外推时水位系数不均匀的影响,使得外推结果更准确。     

核临界安全中子吸收体干涉效应实验研究

简要阐述了干涉效应的原理、铀溶液实验装置的临界测量实验,研究了多组固体中子吸收体在装置容器中的不同位置、不同铀溶液浓度、不同组合情况下的吸收效率,并给出干涉效应。测量结果表明,偏心对称布置的干涉效应为正,偏心非对称布置的干涉效应为负。同时,利用蒙特卡罗程序分别对固体中子吸收体不同布置和组合情况下的中子吸收效率进行了计算分析。计算结果表明,实验测量与理论计算的干涉效应大小、正负的变化趋势相互一致,这表明,利用蒙特卡罗程序计算分析铀溶液系统的中子吸收体的干涉效应是适宜的。     

Analysis of SHE Critical Experiments by Neutronic Design Codes for Experimental Very High Temperature Reactor

Analyses have been performed on various experiments conducted using the Semi-Homogeneous Experimental Assembly (SHE) to examine the accuracy of computer codes employed in the neutronic design of experimental Very High Temperature Reactor (VHTR). The neutronic design codes are DELIGHT-6 to obtain the neutron spectrum of a fuel cell and to produce group constants with burnup utilizing the nuclear data from ENDF-B/IV, CITDEGA to calculate the three-dimensional core performance considering the coupling effect between neutronic and thermohydraulic characteristics, and ANISN-JR and TWOTRAN-II for transport calculation. These codes are examined by the analysis on the integral quantities of effective multiplication factor, neutron flux distribution, burnable poison rod worth and control rod worth. The maximum degrees of disagreement with the relevant experiments are 0.57, 5, 7 and 5%, respectively.     

Theoretical Study on New Bias Factor Methods to Effectively Use Critical Experiments for Improvement of Prediction Accuracy of Neutronic Characteristics

Extended bias factor methods are proposed with two new concepts, the LC method and the PE method, in order to effectively use critical experiments and to enhance the applicability of the bias factor method for the improvement of the prediction accuracy of neutronic characteristics of a target core. Both methods utilize a number of critical experimental results and produce a semifictitious experimental value with them. The LC and PE methods define the semifictitious experimental values by a linear combination of experimental values and the product of exponentiated experimental values, respectively, and the corresponding semifictitious calculation values by those of calculation values. A bias factor is defined by the ratio of the semifictitious experimental value to the semifictitious calculation value in both methods. We formulate how to determine weights for the LC method and exponents for the PE method in order to minimize the variance of the design prediction value obtained by multiplying the design calculation value by the bias factor. From a theoretical comparison of these new methods with the conventional method which utilizes a single experimental result and the generalized bias factor method which was previously proposed to utilize a number of experimental results, it is concluded that the PE method is the most useful method for improving the prediction accuracy. The main advantages of the PE method are summarized as follows. The prediction accuracy is necessarily improved compared with the design calculation value even when experimental results include large experimental errors. This is a special feature that the other methods do not have. The prediction accuracy is most effectively improved by utilizing all the experimental results. From these facts, it can be said that the PE method effectively utilizes all the experimental results and has a possibility to make a full-scale-mockup experiment unnecessary with the use of existing and future benchmark experiments.     

ADS铅冷却剂临界装置堆芯物理设计

为研究加速器驱动次临界反应堆系统(ADS)次临界堆芯与靶的耦合特性,以验证设计方法和计算程序,本文构建了ADS特有的快中子谱、较高能量放大系数及负温度系数的铅冷却剂临界装置堆芯,以用于开展不同富集度燃料特性、不同外源能谱与强度条件、不同实验样品的反应性影响、中子源与堆芯耦合特性等实验研究。确定了燃料元件构造、靶区结构、堆芯布置、反射层结构与价值、安全控制及反应性价值等物理参数,为下一步ADS铅冷却剂临界装置研制及实验研究提供了工程实施依据。     

Development of Steam Separator Performance Analysis Code and Its Validation, (II) Carryover Experiments

Abstract

The purpose of the study is to collect data in order to make models that are applicable to calculate carryover droplets that are generated in and flow out of a steam separator. Various effective tests relevant to separation mechanisms in the separator have been conducted with a full-scale steam separator under atmospheric pressure. Separation behaviors for the top of the riser of the separator and for corrugated-separator were clarified and correlated by the experiment. Distinct patterns about the separation at the corrugated-separator, the separation of discharged droplets by gravity, and the separation of droplets by a screen dryer that is used to dry up the steam were also measured with the facility using a full-scale separator in a vessel simulating the flow area of ATR under the high-pressure and high-temperature condition for various water levels. Each separation data were correlated under the condition of maximum steam and liquid flow rates of 7 and 30.5 kg/s, respectively. Liquid droplets containing a small amount of LiOH at several positions were sampled together with steam by iso-kinetic probes, and the amount of carryover was analyzed in PPT range by the chemical analysis of condensed steam. As a result, basic data for separation mechanisms were obtained, and maximum capacity of the separator was estimated.     

临界计算程序CHMCK-Ⅲ的临界安全基准检验计算

利用4个快中子基准实验数据和MCNP-4B程序的计算结果，对临界计算程序CHMCK-Ⅲ进行了检验计算。从两个程序的计算结果与基准实验测量结果的比较可看出：计算结果在误差允许的范围内符合较好，因此，认为CHMCK-Ⅲ可用于核系统临界计算问题。     

一体化装置实验对RELAP5的评价

针对一体化自然循环试验装置OSU-MASLWR开展的实验,采用系统分析程序RELAP5/MOD3.3进行分析计算。失水事故瞬态计算结果表明,堆芯有足够的冷却,加热元件在整个瞬态过程中温度不断降低。安全壳内出现热分层现象,通过安全壳壁从安全壳到周围水池的热传递速率足以除去堆芯的衰变热。与试验结果相比,程序基本预测了整个瞬态过程中各参数的变化情况。     

DRAGON程序加载WLUP数据库处理钍基燃料的基准题计算验证

为验证DRAGON程序加载WLUP数据库处理钍基燃料问题的可靠性,本文使用DRAGON程序加载WLUP提供的14种WIMSLIB格式核数据库,计算钍基燃料基准问题的k_eff并与实验值进行比对,选择IAEA提供的WIMSD程序计算结果作为比对组。结果表明：DRAGON程序计算结果与WIMSD程序计算结果表现出较好的吻合性,处理轻水慢化钍基燃料时,推荐使用endf68gx数据库,其平均相对偏差为0.18%;处理重水慢化钍基燃料时,推荐使用endf71与jendl3gx数据库,其平均相对偏差为0.81%。因此,使用DRAGON程序加载合适的WLUP数据库计算钍基燃料问题具有一定的可行性。     

FRAPCON-2程序改进和校验

通过改进FRAPCON-2程序中的燃料导热系数模型和裂变气体释放模型,使之能对高燃耗的燃料进行性能分析计算。并利用Halden堆IFA 597.3 ROD8的试验数据对程序进行了验证。结果表明,改进后的程序所计算出的参数（如燃料温度和裂变气体释放份额）均与实测值符合很好,对程序的改进是成功的。     

核反应堆物理确定论计算程序系统NECP软件包的研发与应用

NECP软件包是西安交通大学反应堆物理团队开发的确定论核反应堆物理计算程序系统,软件包包括自主化的NECP-Atlas、Bamboo、X和SARAX程序。NECP软件包经过了大量的验证与确认。数值结果表明,NECP软件包精度高,可满足不同反应堆物理计算需求,具有高度的通用性并实现了对压水堆的高保真建模和计算。目前程序已应用于我国大型压水堆项目、示范快堆项目等重点工程。应用结果表明,NECP软件包已达到甚至优于国际先进核设计程序水平,对我国核电软件自主化和核设计能力提升具有重要的意义。