Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.     

蒙卡-燃耗程序系统及ADS基准题的计算

为对核废物进行嬗变,提出了加速器驱动的次临界系统(ADS)。由于外中子源的存在及中子注量率的各向异性,ADS的核计算和常规反应堆的有较大的差别,原则上需要应用中子输运理论的方法,目前尚无成熟的计算方法与程序。为此,IAEA提出了ADS基准题。基准题分为几个阶段,第一阶段主要致力于ADS的中子性能分析,检验现有核数据库、计算方法和程序的可靠性及不确定性。开发了MC-NT-ORIGEN2程序系统,并对该基准题进行了给定次临界点下的富集度、零燃耗下的径向与轴向的功率分布、反应性空泡效应、外源价值和燃耗的计算,取得了满意的结果。     

蒙卡燃耗计算程序的工程开发研究

分析了反应堆设计中燃耗计算程序的基本功能需求,指出了相关功能在传统确定论中的实现方式并不适用于蒙卡燃耗计算程序,针对蒙卡燃耗计算程序MOI精确组合几何与复杂燃耗链的特点,为MOI程序开发了应用于燃料/可燃毒物燃耗功能的混合燃耗模式,采用更精确的体移动实现调棒临界功能,同时还开发了复杂燃耗链下的再启动功能,初步实现了蒙卡燃耗计算程序MOI在工程上的可用性。     

COCOSYS: Status of development and validation of the German containment code system

For the simulation of severe accident propagation in containments of nuclear power plants it is necessary to assess the efficiency of severe accident measures under conditions as realistic as possible. Therefore the German containment code system COCOSYS is under development and validation at GRS. The main objective is to provide a code system on the basis of mostly mechanistic models for the comprehensive simulation of all relevant processes and plant states during severe accidents in the containment of light water reactors covering the design basis accidents, too. COCOSYS is being used for the identification of possible deficits in plant safety, qualification of the safety reserves of the entire system, assessment of damage-limiting or mitigating accident management measures, support of integral codes in PSA level 2 studies and safety evaluation of new plants.COCOSYS is composed of three main modules, which are separate executable files. These modules are covering thermal hydraulics including hydrogen combustion, fission products mainly aerosols and iodine behaviour, and corium behaviour with molten corium concrete interaction. The communication between these modules is realized via PVM (parallel virtual machine).COCOSYS is subject to an ongoing internal and external validation process. At present this validation process is mainly based on tests being performed in the German ThAI facility. Experiments to be performed in ThAI dealing with hydrogen combustion, recombiner behaviour and aerosol and iodine issues are currently subject of the just started OECD-THAI project. Examples given for the successful validation are the participation in the OECD/NEA ISP-47 and the benchmark for the CCI-2 test in the frame of the OECD-MCCI project.For example COCOSYS has been used in licensing procedure performed for the installation of catalytic recombiners in German nuclear power plants. At present COCOSYS is in use for the licensing process of the new Finnish EPR plant on the industrial side.Improvements and model extensions like pyrolysis processes, direct containment heating and the combined use with CFD models are just ongoing.     

Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor–corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP–ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems’ k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.     

ADS加速器束流瞬变分析程序开发

加速器驱动的反应堆系统(ADS)中次临界堆芯的功率水平依靠强流质子轰击散裂靶产生的中子源来维持,质子束流的不稳定性将对次临界堆的功率水平产生影响,进而对ADS的安全性产生影响.本文研究了ADS系统束流瞬变事故特性,建立了相应的物理数学模型,设计开发出具有较强针对性的用于ADS系统束流瞬变事故仿真软件--SIMULINK-ADS.并选取了典型的束流瞬变工况进行分析,通过与OECD/NEA和FZK Karlsruhe研究成果进行比较,验证了SIMULINK-ADS程序能够有效地计算和分析ADS束流瞬变次临界反应堆堆芯物理及热工响应.     

Development of a thermal–hydraulic system code for simulators based on RELAP5 code

A thermal–hydraulic system code for simulators, RELAPSIM, was developed at NSSE based on RELAP5. The development procedure consists of three major parts. Firstly, time control function was added into the code to meet real-time calculation needs. Secondly, controlled dynamic data communication was improved, so that thermal–hydraulic parameters can be easily modified for further applications. Finally, functions controlling the computation procedure were embedded to achieve a full capability to simulate multiple operations, such as start-up, shutting down or freeze. This paper describes the main features of the new code. The results of code assessment and code application are presented and discussed.     

Experimental validation of the thermal-hydraulic code SACATRI

A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.     

Prediction of MgO volume fraction in an ADS fresh fuel for the scenario code CLASS

Subcritical reactors, also called Accelerator Driven Systems (ADS), are specifically studied for their capacity in transmuting Minor Actinides (MA). Nuclear fuel cycle scenarios involving MA transmutation in ADS are widely researched. The nuclear fuel cycle simulation tool code CLASS (Core Library for Advanced Scenarios Simulations) is dedicated to the inventory evolution calculation induced by a complex nuclear fleet. For managing reactors, the code CLASS includes physic models. Loading models aim to provide the fuel composition at beginning of cycle according to the stocks isotopic composition and the reactors requirements. A cross section predictor aims to provide mean cross sections needed for solving Bateman equations. Physic models are built from reactors calculation set ahead of the scenario calculation. An ADS standard composition at BOC is a mixture of plutonium and MA oxide. The high number of fissile isotopes present in the subcritical core leads to an issue for building an ADS fuel loading model. A high number of isotopic vector at BOC is needed to get an exhaustive simulation set. Also, ADS initial reactivity is adjusted with an inert matrix which induces an additional degree of freedom. The building of an ADS fuel loading model for CLASS requires two steps. For any heavy nuclide composition at beginning of cycle, the core reactivity must be imposed at a subcritical level. Also, the reactivity coefficient evolution should be maintained during the irradiation. In this work, the MgO volume fraction is adjusted to reach the first requirement. The methodology based on a set of reactor simulations and neural network utilization to predict the MgO volume fraction needed to reach a wanted k_eff for any initial composition is presented. Also, a complete neutronic study is done that highlight the effect on MgO on neutronic parameters. Reactor simulations are done with the transport code MCNP6 (Monte Carlo N particle transport code). The ADS geometry is based on the EFIT (European Facility for Industrial-Scale Transmutation) concept. The simulation set is composed of more than 8000 randomized runs from which a neural network has been built. The resulting MgO prediction method allows reaching a k_eff at 0.96 and the distribution standard deviation is around 200 pcm.     

Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.     

Development and Verification of PWR Burnup Code Based on CRAM

<正>A burnup calculation code MCRAM based on CRAM was developed.The Cinder90burnup database was introduced to generate the burnup matrix(Fig.1).The reaction cross sections of some important actinides and fission products were modified by coupling with the MCNP program.The OECD/NEA Takahama-3 PWR fuel     

Development and validation of a real-time SAFT-UT system for inservice inspection of LWRs

A three-year program at the Pacific Northwest Laboratory is underway to move the synthetic aperture focusing technology from the laboratory into the field. This is a report of the first year's activities that concentrated on conducting the needed engineering for a field system. In addition, the data acquisition portion of the system was completed so that data can be collected in the field with subsequent laboratory processing. The field system specifications are given and some SAFT results are shown.     

Development and validation of a real-time SAFT-UT system for inservice inspection of LWRs

A three-year program is in progress at the Pacific Northwest Laboratory to evolve the synthetic aperture focusing technique for ultrasonic testing (SAFT-UT) from the laboratory into the field for inspection of light water reactor components. This paper reports on the second year's activities and highlights work on a time variable gain amplifier, laboratory work to develop tandem SAFT (TSAFT), attempts at reducing processing time, envelope detection techniques, and field trips to Dresden and Vermont Yankee nuclear power plants.     

Improvements and validation of the transient analysis code MOREL for molten salt reactors

The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective.     

The development and validation of a natural circulation analysis code for marine reactors

A code PNCMC (Program for Natural Circulation under Motion Conditions) has been developed for natural circulation simulation of marine reactors. The code is based on one-dimensional two-fluid model in noninertial frame of reference. The body force term in the momentum equation is considered as a time dependent function, which consists of gravity and inertial force induced by three-dimensional ship motion. Staggered mesh, finite volume method, semi-implicit first order upwind scheme and Successive Over Relaxation (SOR) method are used to discretize and solve two-phase mass, momentum and energy equations. Single-phase natural circulation experiments under rolling condition performed in Institute of nuclear and new energy technology of Tsinghua University and two-phase natural circulation experiments under rolling condition performed by Tan and colleagues are used to validate PNCMC. The validation results indicate that PNCMC is capable to investigate the single-phase and two-phase natural circulation under rolling motion.     

压水堆燃耗库的开发与验证

基于最新发布的评价数据库ENDF/B-Ⅷ.0、FENDL-3.1D和EAF-2010开发了一个压水堆(PWR)用燃耗库BURN.SNERLIB,用于ORIGEN-S程序。此数据库由三部分组成:衰变数据、裂变产额数据和截面数据,其格式与ORIGEN-S自带压水堆数据库的格式保持一致。衰变数据选取MF=8文档中的MT=457反应数据进行加工;裂变产额数据共考虑了30种可裂变锕系核素,由特定入射能量下MT=454和MT=459反应数据加工得到;截面数据采用三群结构,首先基于典型压水堆燃料棒栅元在指定燃耗深度下的输运计算获得燃料区域内逐点中子能谱,以此逐点中子谱为权重谱通过NJOY程序将ENDF/B-Ⅷ.0等评价库中的连续截面制作成精细群截面,对精细群截面进行并群计算生成三群截面。利用OECD/NEA公布的压水堆基准题进行了验证,验证了此方法加工ORIGEN-S燃耗库的正确性。分析结果表明,对于某些燃耗计算重要核素,如238 Pu等,基于最新评价库开发的数据库比自带库的计算结果更接近于实验值,提升了ORIGEN-S的计算精度。     

Development of a consensus standard for verification and validation of nuclear system thermal-fluids software

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.     

基于切比雪夫有理近似法的燃耗程序开发

基于切比雪夫有理近似法(CRAM)对燃耗方程进行求解,采用EAF数据库,开发了燃耗程序ABURN。计算了聚变堆第一壁活化例题和UO2燃料燃耗例题,并将ABURN程序的计算结果与欧洲活化程序FISPACT进行对比。结果表明,ABURN程序可达到FISPACT程序同等精度,并且由于采用了CRAM,程序在燃耗步设置方面具有高度的灵活性,初步验证了ABURN程序的可用性与准确性。     

非破坏性燃耗测量方法综述

简述了国内外研究和应用的各种非破坏性燃耗测量技术和方法,并介绍了各国所研制的非破坏性燃耗测量装置及其用途和特点.