共查询到17条相似文献,搜索用时 234 毫秒
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COSINE软件包堆芯物理分析程序CORE开发与初步测试验证 总被引:1,自引:1,他引:0
堆芯物理分析程序CORE是1个少群、一维、二维、三维稳态节块法程序,用于压水堆堆芯设计和分析。COSINE软件包是大型压水堆国家重大专项软件自主化课题中的一部分,CORE是COSINE软件包的1个子程序系统,CORE第1版采用节块展开法(NEM)进行二维、三维扩散计算,采用差分法进行一维扩散计算,截面处理采用插值表的方式,燃耗计算采用带预估修正的宏观燃耗计算方法,精细功率重构采用调制方法。目前CORE的核心模块已完成,并进行了初步测试验证,结果表明其扩散求解模块基本满足功能和精度要求。 相似文献
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COSINE软件包是国家核电技术有限公司北京软件技术中心自主研发的堆芯设计和安全分析一体化软件包,安注箱是该软件包用于压水堆核电厂主回路分析的重要设备。本文介绍了基于NewtonRaphson算法建立的安注箱模型,该模型采用半隐式离散方法,程序的四个守恒方程采用有限体积法,其方程分别为:不凝气体质量守恒方程,液相质量守恒方程、气相动量守恒方程以及气相能量守恒方程。守恒方程考虑惯性力,壁面摩擦,局部阻力损失以及重力等因素影响。利用COSINE安注箱模型和参考程序RELAP5计算结果对比,两者计算结果误差较小,吻合较好;结果表明:COSINE安注箱模型能正确模拟安注箱,个别情况下COSINE安注箱模块更具稳定性和收敛性。 相似文献
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cosFlow软件是国家电投集团科学技术研究院有限公司核电软件技术中心自主研发的堆芯设计和安全分析一体化软件包COSINE中的热工水力与安全分析软件。主泵是cosFlow软件中用于压水堆核电厂主回路分析的重要设备。本文介绍了基于动力学原理的主泵模型,该模型采用惯量方程计算泵的角速度,采用泵的特性曲线得到泵的扬程值。本文搭建了泵的简单算例和全失流事故算例,利用COSINE主泵模型和参考程序RELAP计算结果对比,两者计算结果趋势一致,吻合较好;结果表明:COSINE主泵模型能正确模拟主泵,COSINE主泵模型能够很好的进行核电厂设计和安全分析。 相似文献
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安全软件验证与确认中的单元模块测试技术 总被引:1,自引:0,他引:1
核动力厂保护系统实现数字化必须解决的一项关键技术是如何完成安全软件的验证与确认(V&V),以证明和确认执行安全功能的软件自身的安全性和可靠性。软件单元测试是V&V过程中的重要环节,主要目的是验证和确认软件代码开发过程中,软件的设计转变为软件代码是适当、正确和完整的。本工作初步研究了安全软件的单元模块测试技术,着重讨论如何保证测试的完整性、建立测试环境、建立测试用例及实施单元模块测试等方面,并以某数字化保护系统安全软件单元模块测试实例说明单元模块测试的具体过程。 相似文献
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为能更加准确地模拟典型压水堆中强烈的物理-热工耦合现象,研制了压水堆堆芯物理 热工耦合计算软件ARMcc。其中物理计算模块基于四阶节块展开法(NEM)和格林函数节块法(NGFM),热工计算模块基于一维的单相单通道换热模型和一维圆柱导热计算模型,在程序中采用有限体积法和有限差分法求解一维圆柱导热模型。基于典型压水堆基准题NEACRP-L-335对程序的稳态耦合计算能力进行了验证,程序计算的堆芯关键参数如临界硼浓度、堆芯多普勒温度等参数与参考结果符合良好,临界硼浓度与参考结果的相对偏差均小于0.5%。另外研究4种计算模式对模拟堆芯物理-热工耦合过程的影响,选择PARCS程序计算结果为对比,发现NGFM+DIF模式能更加准确地模拟堆芯燃料多普勒温度和堆芯功率分布;NGFM+VOL模式能更加准确地模拟临界硼浓度;NEM+VOL模式能更加准确地模拟堆芯燃料最高温度。 相似文献
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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。 相似文献
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Verification and validation benchmarks 总被引:3,自引:0,他引:3
Verification and validation (V&V) are the primary means to assess the accuracy and reliability of computational simulations. V&V methods and procedures have fundamentally improved the credibility of simulations in several high-consequence fields, such as nuclear reactor safety, underground nuclear waste storage, and nuclear weapon safety. Although the terminology is not uniform across engineering disciplines, code verification deals with assessing the reliability of the software coding, and solution verification deals with assessing the numerical accuracy of the solution to a computational model. Validation addresses the physics modeling accuracy of a computational simulation by comparing the computational results with experimental data. Code verification benchmarks and validation benchmarks have been constructed for a number of years in every field of computational simulation. However, no comprehensive guidelines have been proposed for the construction and use of V&V benchmarks. For example, the field of nuclear reactor safety has not focused on code verification benchmarks, but it has placed great emphasis on developing validation benchmarks. Many of these validation benchmarks are closely related to the operations of actual reactors at near-safety-critical conditions, as opposed to being more fundamental-physics benchmarks. This paper presents recommendations for the effective design and use of code verification benchmarks based on manufactured solutions, classical analytical solutions, and highly accurate numerical solutions. In addition, this paper presents recommendations for the design and use of validation benchmarks, highlighting the careful design of building-block experiments, the estimation of experimental measurement uncertainty for both inputs and outputs to the code, validation metrics, and the role of model calibration in validation. It is argued that the understanding of predictive capability of a computational model is built on the level of achievement in V&V activities, how closely related the V&V benchmarks are to the actual application of interest, and the quantification of uncertainties related to the application of interest. 相似文献
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VSOP程序广泛用于球床式高温气冷堆的工程设计,需对VSOP程序进行验证与确认。针对相同的高温气冷堆堆芯定义,比较了VSOP程序和蒙特卡罗程序采用特殊形状的重复结构来模拟随机分布的球床堆芯的建模结果。对VSOP模型中的侧反射层孔道进行均匀化处理,对球床顶锥和底锥的几何进行等效处理,用蒙特卡罗模型详细比较了近似处理为有效增殖因数keff带来的偏差。结果表明,VSOP模型中不同的近似处理方法会带来不同的偏差。但最终VSOP模型与最精细的蒙特卡罗模型在有效增殖因数方面差别不大,进一步说明VSOP模型的可用性和合理性。当然,VSOP程序和模型的验证还需要进一步深入研究。 相似文献
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COSINE多相场子通道分析程序基于两流体三相子通道守恒方程,在气液两相的基础上,单独考虑了液滴相的行为,并通过考虑通道间的交混,提高了对压水堆压力容器内的热工水力学现象分析能力及大破口事故的计算能力。本研究介绍了程序的基本模型及求解方法,选取代表性算例及实验工况进行建模计算,验证多相场子通道程序的计算能力。计算结果表明:程序可以对多通道热工水力现象进行模拟计算,计算结果与理论分析相符,程序可以精确模拟堆芯交混及再淹没工况,计算结果与实验数据具有良好的一致性,COSINE多相场子通道程序具备对压力容器内热工水力工况的计算能力。 相似文献
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Edwin A. Harvego Richard R. Schultz Ryan L. Crane 《Nuclear Engineering and Design》2011,241(12):4691-4696
With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V&V) of software used to calculate the thermal–hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V&V 30 Committee, under the jurisdiction of the V&V Standards Committee, to develop a consensus standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V&V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. Although software verification will be an important and necessary part of the standard, much of the initial effort of the committee will be focused on the validation of existing software and new models that could be used in the licensing process. In this framework, the Standard should conform to Nuclear Regulatory Commission (NRC) and other regulatory practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and other pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the Standard should be consistent with applicable sections of ASME NQA-1-2008 “Quality Assurance Requirements for Nuclear Facility Applications (QA)”. This paper describes the general requirements for the proposed V&V 30 Standard, which includes: (a) applicable NRC and other regulatory requirements for defining the operational and accident domain of a nuclear system that must be considered if the system is to be licensed, (b) the corresponding calculation domain of the software that should encompass the nuclear operational and accident domain to be used to study the system behavior for licensing purposes, (c) the definition of the scaled experimental data set required to provide the basis for validating the software, (d) the ensemble of experimental data sets required to populate the validation matrix for the software in question, and (e) the practices and procedures to be used when applying a validation standard. Although this initial effort will focus on software for licensing of High-Temperature Gas-Cooled Reactors, it is anticipated that the practices and procedures developed for this Standard can eventually be extended to other nuclear and non-nuclear applications. 相似文献
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The reactor protection system (RPS) used in the 10 MW high-temperature gas-cooled reactor is the first digital RPS designed and operated in China. In order to ensure its safety and reliability and to reduce the development risk and cost, some measures had to be taken. The measures adopted in the development process include the architecture of defense-in-depth, commercial grade hardware, prototype development model, separation of safety class software and non-safety class software, deterministic behavior of safety software, etc. The measures adopted in the verification and validation (V&V) process include effective dedication on the commercial grade hardware, emphasis on the assessment of the requirements and specifications, emphasis on the demonstration and testing, thorough testing for the safety function, long period demonstration operation, application of automatic test system to improve the efficiency of V&V processes, etc. As a result, this first digital RPS has passed the safety assessment of the National Nuclear Safety Authority. Its performance and safety are proven to be confident and assuring through the demonstration and testing. Thus, the design and V&V process of the first digital protection system in China was successful. 相似文献