Pub Date : 2024-09-27DOI: 10.1016/j.anucene.2024.110947
Targets are irradiated in high-flux reactors to produce transplutonium isotopes. Neutron environment of the target is crucial for the production efficiency of transplutonium isotopes. To improve the production efficiency of transplutonium isotopes, it is necessary to research the optimization design of target. Taking the production of Californium-252 as an example, this study analyzed the impact of self-shielding effect in targets on the yield of transplutonium isotope based on the High Flux Isotope Reactor (HFIR) and High-Flux Fast Reactor (HFFR). The self-shielding effect leads to the hardening of the neutron spectrum inside the target and significantly reduces the conversion rate of nuclides. After conducting a refined energy spectrum analysis, we proposed a layered target design method based on the Genetic Algorithm (GA). To reduce computational costs, we propose a fixed source-burnup coupling approximate calculation method, which can avoid tedious burnup calculation and provide optimization direction. Using this method, we designed an optimal layered target scheme. Compared with non-layered target, the production efficiency of Cf-252 was increased by approximately 4.1 times. This study provides technical support for energy spectrum analysis and target design in producing transplutonium isotopes.
{"title":"Layered target design method for global spectrum optimization of radioisotope production","authors":"","doi":"10.1016/j.anucene.2024.110947","DOIUrl":"10.1016/j.anucene.2024.110947","url":null,"abstract":"<div><div>Targets are irradiated in high-flux reactors to produce transplutonium isotopes. Neutron environment of the target is crucial for the production efficiency of transplutonium isotopes. To improve the production efficiency of transplutonium isotopes, it is necessary to research the optimization design of target. Taking the production of Californium-252 as an example, this study analyzed the impact of self-shielding effect in targets on the yield of transplutonium isotope based on the High Flux Isotope Reactor (HFIR) and High-Flux Fast Reactor (HFFR). The self-shielding effect leads to the hardening of the neutron spectrum inside the target and significantly reduces the conversion rate of nuclides. After conducting a refined energy spectrum analysis, we proposed a layered target design method based on the Genetic Algorithm (GA). To reduce computational costs, we propose a fixed source-burnup coupling approximate calculation method, which can avoid tedious burnup calculation and provide optimization direction. Using this method, we designed an optimal layered target scheme. Compared with non-layered target, the production efficiency of Cf-252 was increased by approximately 4.1 times. This study provides technical support for energy spectrum analysis and target design in producing transplutonium isotopes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323277","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-26DOI: 10.1016/j.anucene.2024.110917
Griffin is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based reactor physics application for multiphysics simulations of advanced reactor designs jointly developed by Idaho National Laboratory and Argonne National Laboratory. This paper summarizes the motivation, significance, architecture, design, and features of Griffin. Griffin offers flexible and extensible features to address the challenges associated with advanced reactor designs. These features range from fundamental particle transport to specific reactor physics tasks. The features cover a wide range including on-the-fly and traditional two-step cross-section generation methods, steady-state and transient transport solvers suitable for both heterogeneous and homogeneous models, high-fidelity depletion where thousands of isotopes can be tracked and low-fidelity depletion characterized by burnup, etc. The most fundamental aspect that sets Griffin apart from other reactor analysis codes is that it is developed based on the MOOSE framework. A modular development approach is strongly enforced, with multiphysics being an essential element considered since the beginning of Griffin’s development. Griffin links various MOOSE physics modules and couples to other MOOSE-based applications and non-MOOSE-based applications for multiphyiscs simulations. Griffin includes three modules: ISOXML for preparing and managing multigroup cross sections, radiation transport for solving the neutron transport equation, and reactor analysis for user-oriented reactor physics analysis functionalities. Griffin uses various finite element methods for spatial discretization, multigroup approximation for energy discretization and discrete ordinates method, spherical harmonics expansion method, and diffusion approximation for streaming direction discretization to solve the neutron transport equation. Griffin’s flexibility is evidenced through Griffin’s various applications to fast reactor, high-temperature reactor, pebble bed reactor, molten salt reactor, and microreactor designs. Griffin development follows the software quality assurance procedure for MOOSE-based applications and with software requirements consistent with the ASME NQA-1 standard. Griffin has been adopted into the reactor analysis system for the U.S. NRC and is in use at U.S. companies, universities and national laboratories.
{"title":"Griffin: A MOOSE-based reactor physics application for multiphysics simulation of advanced nuclear reactors","authors":"","doi":"10.1016/j.anucene.2024.110917","DOIUrl":"10.1016/j.anucene.2024.110917","url":null,"abstract":"<div><div>Griffin is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based reactor physics application for multiphysics simulations of advanced reactor designs jointly developed by Idaho National Laboratory and Argonne National Laboratory. This paper summarizes the motivation, significance, architecture, design, and features of Griffin. Griffin offers flexible and extensible features to address the challenges associated with advanced reactor designs. These features range from fundamental particle transport to specific reactor physics tasks. The features cover a wide range including on-the-fly and traditional two-step cross-section generation methods, steady-state and transient transport solvers suitable for both heterogeneous and homogeneous models, high-fidelity depletion where thousands of isotopes can be tracked and low-fidelity depletion characterized by burnup, etc. The most fundamental aspect that sets Griffin apart from other reactor analysis codes is that it is developed based on the MOOSE framework. A modular development approach is strongly enforced, with multiphysics being an essential element considered since the beginning of Griffin’s development. Griffin links various MOOSE physics modules and couples to other MOOSE-based applications and non-MOOSE-based applications for multiphyiscs simulations. Griffin includes three modules: ISOXML for preparing and managing multigroup cross sections, radiation transport for solving the neutron transport equation, and reactor analysis for user-oriented reactor physics analysis functionalities. Griffin uses various finite element methods for spatial discretization, multigroup approximation for energy discretization and discrete ordinates method, spherical harmonics expansion method, and diffusion approximation for streaming direction discretization to solve the neutron transport equation. Griffin’s flexibility is evidenced through Griffin’s various applications to fast reactor, high-temperature reactor, pebble bed reactor, molten salt reactor, and microreactor designs. Griffin development follows the software quality assurance procedure for MOOSE-based applications and with software requirements consistent with the ASME NQA-1 standard. Griffin has been adopted into the reactor analysis system for the U.S. NRC and is in use at U.S. companies, universities and national laboratories.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323273","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-26DOI: 10.1016/j.anucene.2024.110942
Nuclear thermal propulsion (NTP) is a promising advanced technology which has attracted wide attention in recent years. The reactor core is an essential component of an NTP system and the corresponding thermal–hydraulic analysis is necessary. In this study, the porous medium approach was applied to the simulation of a two-pass NTP reactor core which consists of the porous prismatic cermet fuel elements. The thermodynamic property models of hydrogen and the fuel element materials were implemented, as well as the empirical correlations of the heat transfer coefficient and the friction factor. The three-dimensional simulation of a single fuel element was carried out and the results were compared against another code. The code-to-code comparison verified the applicability of the porous medium approach. The three-dimensional model of the two-pass NTP reactor core was established and the steady-state simulation was carried out. The distribution patterns of the parameters are determined by the thermal–hydraulic characteristics of the reactor core, including the nonuniform heat release, contact heat conduction and folded-flow scheme. The full-core heat-flow adaptability analysis is realized, which provides a reference for the thermal–hydraulic safety analysis of the NTP reactor.
{"title":"Steady-state thermal–hydraulic analysis of an NTP reactor core based on the porous medium approach","authors":"","doi":"10.1016/j.anucene.2024.110942","DOIUrl":"10.1016/j.anucene.2024.110942","url":null,"abstract":"<div><div>Nuclear thermal propulsion (NTP) is a promising advanced technology which has attracted wide attention in recent years. The reactor core is an essential component of an NTP system and the corresponding thermal–hydraulic analysis is necessary. In this study, the porous medium approach was applied to the simulation of a two-pass NTP reactor core which consists of the porous prismatic cermet fuel elements. The thermodynamic property models of hydrogen and the fuel element materials were implemented, as well as the empirical correlations of the heat transfer coefficient and the friction factor. The three-dimensional simulation of a single fuel element was carried out and the results were compared against another code. The code-to-code comparison verified the applicability of the porous medium approach. The three-dimensional model of the two-pass NTP reactor core was established and the steady-state simulation was carried out. The distribution patterns of the parameters are determined by the thermal–hydraulic characteristics of the reactor core, including the nonuniform heat release, contact heat conduction and folded-flow scheme. The full-core heat-flow adaptability analysis is realized, which provides a reference for the thermal–hydraulic safety analysis of the NTP reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323382","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-26DOI: 10.1016/j.anucene.2024.110941
In this paper, low-temperature experiments are carried out on the visualized experimental device to study the melting behavior of plate-type fuel in severe accidents of the reactor. In the experiments, the plate-type fuel with different sizes made of nickel–chromium alloy, zinc and aluminum was used to carry out the visualized experiments in air, argon, and vacuum environment. It was found that both the size of the plate and the experimental environment have a significant influence on the melting behavior in this study. And the temperature distribution, melting behavior characteristic, the key parameters such as blistering position, blistering size, breaking position and breaking size were also obtained. Based on the experimental data, the physical phenomena and processes related to the blistering and melting of the fuel plates are analyzed in this paper, which provides experimental data support for the development of analysis model and formulating perfect mitigation strategies for severe accidents.
{"title":"Experimental study on the plate-type fuel melting behavior based on alternative materials","authors":"","doi":"10.1016/j.anucene.2024.110941","DOIUrl":"10.1016/j.anucene.2024.110941","url":null,"abstract":"<div><div>In this paper, low-temperature experiments are carried out on the visualized experimental device to study the melting behavior of plate-type fuel in severe accidents of the reactor. In the experiments, the plate-type fuel with different sizes made of nickel–chromium alloy, zinc and aluminum was used to carry out the visualized experiments in air, argon, and vacuum environment. It was found that both the size of the plate and the experimental environment have a significant influence on the melting behavior in this study. And the temperature distribution, melting behavior characteristic, the key parameters such as blistering position, blistering size, breaking position and breaking size were also obtained. Based on the experimental data, the physical phenomena and processes related to the blistering and melting of the fuel plates are analyzed in this paper, which provides experimental data support for the development of analysis model and formulating perfect mitigation strategies for severe accidents.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323193","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-26DOI: 10.1016/j.anucene.2024.110943
Due to the uncertainty of computational results and the lack of interpretability of models in solving physical field equations in current deep learning, this paper designs a convolutional neural network that can be used to solve the neutron diffusion spatiotemporal kinetics equation in polar and cylindrical coordinate systems. This algorithm directly utilizes the macroscopic cross-section of the material without using the lattice homogenization method, replaces the finite volume method with the extended matrices, and solves the extended matrices using the convolutional kernels instead of the iterative algorithms. Taking the simplified Tsinghua High Flux Reactor (THFR) as an example, the feasibility of the algorithm is verified on the PyTorch platform and compared with the calculation results of the source iteration method running on the GPU. The calculation results show that when the number of grids in the radial and axial sections of the simplified THFR model is 804,600 and 3,576,000, respectively, and the algorithm is iterated 3000 times, the normalized power of the convolutional neural network and the source iteration method converges to 10−10, and the maximum point by point error of the neutron flux density of the above two algorithms converges to 10−5. The computational time consumed by the convolutional neural network is approximately 880.64 s and 3729.62 s, which reduces the computational time by 4.66% and 5.05% compared to the GPU parallel accelerated source iteration method, and the former consumes 43.75% less memory compared to the latter. The convolutional neural network is mainly used as the virtual physics engine for the THFR digital twin system, in addition to solving the neutron diffusion spatiotemporal kinetics equation and further improving computational speed. The algorithm directly utilizes the neutron macroscopic cross-section of the material to calculate the neutron flux density distribution without using the lattice homogenization, providing theoretical guidance and algorithm support for developing the high-precision multi-physical field coupling model.
{"title":"Research on the high-performance computing method for the neutron diffusion spatiotemporal kinetics equation based on the convolutional neural network","authors":"","doi":"10.1016/j.anucene.2024.110943","DOIUrl":"10.1016/j.anucene.2024.110943","url":null,"abstract":"<div><div>Due to the uncertainty of computational results and the lack of interpretability of models in solving physical field equations in current deep learning, this paper designs a convolutional neural network that can be used to solve the neutron diffusion spatiotemporal kinetics equation in polar and cylindrical coordinate systems. This algorithm directly utilizes the macroscopic cross-section of the material without using the lattice homogenization method, replaces the finite volume method with the extended matrices, and solves the extended matrices using the convolutional kernels instead of the iterative algorithms. Taking the simplified Tsinghua High Flux Reactor (THFR) as an example, the feasibility of the algorithm is verified on the PyTorch platform and compared with the calculation results of the source iteration method running on the GPU. The calculation results show that when the number of grids in the radial and axial sections of the simplified THFR model is 804,600 and 3,576,000, respectively, and the algorithm is iterated 3000 times, the normalized power of the convolutional neural network and the source iteration method converges to 10<sup>−10</sup>, and the maximum point by point error of the neutron flux density of the above two algorithms converges to 10<sup>−5</sup>. The computational time consumed by the convolutional neural network is approximately 880.64 s and 3729.62 s, which reduces the computational time by 4.66% and 5.05% compared to the GPU parallel accelerated source iteration method, and the former consumes 43.75% less memory compared to the latter. The convolutional neural network is mainly used as the virtual physics engine for the THFR digital twin system, in addition to solving the neutron diffusion spatiotemporal kinetics equation and further improving computational speed. The algorithm directly utilizes the neutron macroscopic cross-section of the material to calculate the neutron flux density distribution without using the lattice homogenization, providing theoretical guidance and algorithm support for developing the high-precision multi-physical field coupling model.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142323276","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-25DOI: 10.1016/j.anucene.2024.110932
CRUD depositions on fuel cladding are the main cause of power shift and localized corrosion in nuclear power plants. This paper is the second of a three-part study concerning the deposition of corrosion products and its related safety issues. In this paper, analytical modules are proposed to predict CRUD growth and internal heat and mass transfer. CRUD growth depends on dynamic balance between corrosion product deposition, flow erosion and chemical equilibrium. In the multi-module iteration, the CRUD thickness is updated first followed by internal temperature and concentration fields. Temperature affects the chemical equilibrium, deposition and erosion equilibrium on CRUD surfaces. The accuracy and reliability of the coupling method are verified by experimental results. The difference of effective thermal conductivity between previous experimental results and calculation results is less than 0.4384 W/(m × K) and the cladding temperature relative error between WALT Loop results and calculation results is less than 1 %. The influences of operation conditions are evaluated. Coolant with lower pH reduces corrosion product solubility leading to high CRUD thickness. The main source of CRUD growth is from soluble precipitation, because CRUD depositions formed from soluble precipitation are thicker than those from the insoluble particles of the same concentration. High heat flux increases CRUD growth, internal wick boiling and boron hideout. Hydrogen in reactor application range has a minimal meaningful effect on CRUD growth, wick boiling and boron hideout. This study provides a precise method for further understanding CRUD growth and its internal multi-physical phenomena to alleviate CRUD-related safety issues.
{"title":"Pressurized water reactor fuel corrosion-related unidentified deposit and its related safety issues – II. Corrosion product deposition and heat transfer modeling","authors":"","doi":"10.1016/j.anucene.2024.110932","DOIUrl":"10.1016/j.anucene.2024.110932","url":null,"abstract":"<div><div>CRUD depositions on fuel cladding are the main cause of power shift and localized corrosion in nuclear power plants. This paper is the second of a three-part study concerning the deposition of corrosion products and its related safety issues. In this paper, analytical modules are proposed to predict CRUD growth and internal heat and mass transfer. CRUD growth depends on dynamic balance between corrosion product deposition, flow erosion and chemical equilibrium. In the multi-module iteration, the CRUD thickness is updated first followed by internal temperature and concentration fields. Temperature affects the chemical equilibrium, deposition and erosion equilibrium on CRUD surfaces. The accuracy and reliability of the coupling method are verified by experimental results. The difference of effective thermal conductivity between previous experimental results and calculation results is less than 0.4384 W/(m × K) and the cladding temperature relative error between WALT Loop results and calculation results is less than 1 %. The influences of operation conditions are evaluated. Coolant with lower pH reduces corrosion product solubility leading to high CRUD thickness. The main source of CRUD growth is from soluble precipitation, because CRUD depositions formed from soluble precipitation are thicker than those from the insoluble particles of the same concentration. High heat flux increases CRUD growth, internal wick boiling and boron hideout. Hydrogen in reactor application range has a minimal meaningful effect on CRUD growth, wick boiling and boron hideout. This study provides a precise method for further understanding CRUD growth and its internal multi-physical phenomena to alleviate CRUD-related safety issues.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142319015","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-25DOI: 10.1016/j.anucene.2024.110948
ATHENA (Advanced Thermal-Hydraulic Experiment for Nuclear Applications) is a large multipurpose pool-type lead-cooled facility under construction at the Mioveni site in Romania. It has been identified by the FALCON (Fostering ALfred CONstruction) Consortium to characterize large to full-scale ALFRED components, to conduct integral tests, and to investigate the main thermal–hydraulic phenomena inherent in pool-type systems. ATHENA is representative of ALFRED in terms of the difference in height of the thermal barycenters of the heat source and heat sink, i.e., 3.3 m, in order to reproduce the buoyancy forces in the system. Similar to ALFRED’s design, ATHENA minimizes thermal stratification within the main vessel even under natural circulation conditions, through an internal structure referred to as “barrel”. This structure directs the fluid flow towards the main vessel, preventing fluid stagnation near the vessel itself. The paper initially provides a steady-state thermal–hydraulic characterization of the facility, including details of the numerical model developed using the RELAP5/Mod3.3 thermal–hydraulic code. Then, focus is given to the transient analysis considering as a reference scenario a Loss-of-Heat-Sink (LOHS) accidental transient. In this scenario, the Main Circulation Pump (MCP) is assumed to remain operational while the Core Simulator (CS) is deactivated once the lead temperature at the Main Heat Exchanger (MHX) outlet reaches a predefined threshold. A sensitivity analysis is conducted with set points of 430 °C, 450 °C, 470 °C, and 490 °C, assessing the system’s response following MHX isolation from the secondary loop. The study evaluates the impact of different CS deactivation set points on reactor SCRAM delay (reducing CS power to a level representative of decay heat) as well as on system maximum and minimum temperatures.
ATHENA（核应用先进热工水力实验）是一个大型多用途池式铅冷设施，正在罗马尼亚米奥韦尼建造。FALCON（促进 ALFRED 建设）财团已确定将其用于鉴定大型乃至全尺寸 ALFRED 组件，进行整体试验，并研究池式系统固有的主要热-水力现象。为了再现系统中的浮力，ATHENA 在热源和散热器的热原点高度差（即 3.3 米）方面具有 ALFRED 的代表性。与 "阿尔弗雷德 "号的设计类似，即使在自然循环条件下，"阿塔纳 "号也能通过一种被称为 "桶 "的内部结构，最大限度地减少主容器内的热分层现象。这种结构将流体流向主容器，防止容器附近的流体停滞。本文首先介绍了该设施的稳态热工水力特性，包括使用 RELAP5/Mod3.3 热工水力代码开发的数值模型的细节。然后，重点介绍瞬态分析，将热沉损失（LOHS）事故瞬态作为参考情景。在这种情况下，假定主循环泵（MCP）保持运行，而一旦主热交换器（MHX）出口处的导线温度达到预定阈值，堆芯模拟器（CS）就会停用。对 430 ℃、450 ℃、470 ℃ 和 490 ℃ 设定点进行敏感性分析，评估 MHX 与二次回路隔离后的系统响应。研究评估了不同的希尔思停用设定点对反应堆 SCRAM 延迟（将希尔思功率降至衰变热量的代表水平）以及系统最高和最低温度的影响。
{"title":"Loss-of-heat-sink transient simulation with RELAP5/Mod3.3 code for the ATHENA facility","authors":"","doi":"10.1016/j.anucene.2024.110948","DOIUrl":"10.1016/j.anucene.2024.110948","url":null,"abstract":"<div><div>ATHENA (Advanced Thermal-Hydraulic Experiment for Nuclear Applications) is a large multipurpose pool-type lead-cooled facility under construction at the Mioveni site in Romania. It has been identified by the FALCON (Fostering ALfred CONstruction) Consortium to characterize large to full-scale ALFRED components, to conduct integral tests, and to investigate the main thermal–hydraulic phenomena inherent in pool-type systems. ATHENA is representative of ALFRED in terms of the difference in height of the thermal barycenters of the heat source and heat sink, i.e., 3.3 m, in order to reproduce the buoyancy forces in the system. Similar to ALFRED’s design, ATHENA minimizes thermal stratification within the main vessel even under natural circulation conditions, through an internal structure referred to as “barrel”. This structure directs the fluid flow towards the main vessel, preventing fluid stagnation near the vessel itself. The paper initially provides a steady-state thermal–hydraulic characterization of the facility, including details of the numerical model developed using the RELAP5/Mod3.3 thermal–hydraulic code. Then, focus is given to the transient analysis considering as a reference scenario a Loss-of-Heat-Sink (LOHS) accidental transient. In this scenario, the Main Circulation Pump (MCP) is assumed to remain operational while the Core Simulator (CS) is deactivated once the lead temperature at the Main Heat Exchanger (MHX) outlet reaches a predefined threshold. A sensitivity analysis is conducted with set points of 430 °C, 450 °C, 470 °C, and 490 °C, assessing the system’s response following MHX isolation from the secondary loop. The study evaluates the impact of different CS deactivation set points on reactor SCRAM delay (reducing CS power to a level representative of decay heat) as well as on system maximum and minimum temperatures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142319016","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-24DOI: 10.1016/j.anucene.2024.110933
The MELCOR code as a severe accident simulation tool does not have the capability to capture the quench process of a debris bed which may form in the wet cavity during a severe accident of light water reactors (LWRs). Although the coupled MELCOR/COCOMO simulation could overcome the limitation (Chen et al., 2022), the calculation time was explosively escalated due to mechanistic modeling of debris bed thermal-hydraulics in COCOMO. To suppress the computational cost, a surrogate model (SM) was developed in our previous study (Wang et al., 2023), and its coupling with MELCOR could realize a quick estimation of the quench process of one-dimensional debris beds. The present study is an extension of the previous work, aiming at the development of a new surrogate model for the quench process of two-dimensional conical debris beds. The new surrogate model (SM) was based on artificial neural networks (ANNs) and trained by the database from COCOMO calculations of various conical debris beds quenched in the reactor cavity of a Nordic boiling water reactor (BWR). The MELCOR was then coupled with the new SM to simulate a postulated station blackout (SBO) scenario in the BWR. The results show that the coupled MELCOR/SM simulation could provide similar ex-vessel debris bed quench period and containment pressure/temperature trends as the coupled MELCOR/COCOMO. Compared with the MELCOR standalone calculation, the coupled calculations predicted earlier points of time for water pool saturation and containment venting, since the heat transfer from conical debris bed to water pool is faster in the coupled simulations.
{"title":"Coupling of MELCOR with surrogate model for quench estimation of conical debris beds","authors":"","doi":"10.1016/j.anucene.2024.110933","DOIUrl":"10.1016/j.anucene.2024.110933","url":null,"abstract":"<div><div>The MELCOR code as a severe accident simulation tool does not have the capability to capture the quench process of a debris bed which may form in the wet cavity during a severe accident of light water reactors (LWRs). Although the coupled MELCOR/COCOMO simulation could overcome the limitation (Chen et al., 2022), the calculation time was explosively escalated due to mechanistic modeling of debris bed thermal-hydraulics in COCOMO. To suppress the computational cost, a surrogate model (SM) was developed in our previous study (Wang et al., 2023), and its coupling with MELCOR could realize a quick estimation of the quench process of one-dimensional debris beds. The present study is an extension of the previous work, aiming at the development of a new surrogate model for the quench process of two-dimensional conical debris beds. The new surrogate model (SM) was based on artificial neural networks (ANNs) and trained by the database from COCOMO calculations of various conical debris beds quenched in the reactor cavity of a Nordic boiling water reactor (BWR). The MELCOR was then coupled with the new SM to simulate a postulated station blackout (SBO) scenario in the BWR. The results show that the coupled MELCOR/SM simulation could provide similar ex-vessel debris bed quench period and containment pressure/temperature trends as the coupled MELCOR/COCOMO. Compared with the MELCOR standalone calculation, the coupled calculations predicted earlier points of time for water pool saturation and containment venting, since the heat transfer from conical debris bed to water pool is faster in the coupled simulations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0306454924005966/pdfft?md5=30a8d621b7bdd294c65c323d9f6699a8&pid=1-s2.0-S0306454924005966-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142311266","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-24DOI: 10.1016/j.anucene.2024.110935
The onset of nucleate boiling (ONB), which marks the emergence of nucleate boiling, is an important transition point in the boiling curve. For exploring the influence of geometric and thermodynamic parameters on ONB in rectangular narrow channels, a detailed experimental study is conducted to investigate ONB under wide range of parameters. The experimental parameters range is pressure of 0.1–5.5 MPa, mass flux of 200–2000 kg/m2s, inlet subcooling of 10–150 K. According to the experimental results, the location of ONB is identified based on the axial distribution of wall temperature, and the influence of various parameters on ONB in narrow rectangular channels is analyzed. It is found that heat flux, pressure, mass flux, and the gap size of the channel have a significant impact on ONB. By comparing the computed results of existing correlations, it is evident that there is a deviation, which can be attributed to the narrow range of experimental parameters in previous studies. Finally, a new ONB model is developed based on basic equations proposed by Hsu and the distribution of liquid temperature, taking into account the influence of mass flux and the enhanced heat transfer results from surrounding bubbles to correct the liquid temperature. The new correlation accurately describes the impact of each parameter and is in good agreement with the current experimental results.
{"title":"Experimental study on onset of nucleate boiling in wide-ranged parameters for narrow rectangular channels","authors":"","doi":"10.1016/j.anucene.2024.110935","DOIUrl":"10.1016/j.anucene.2024.110935","url":null,"abstract":"<div><div>The onset of nucleate boiling (ONB), which marks the emergence of nucleate boiling, is an important transition point in the boiling curve. For exploring the influence of geometric and thermodynamic parameters on ONB in rectangular narrow channels, a detailed experimental study is conducted to investigate ONB under wide range of parameters. The experimental parameters range is pressure of 0.1–5.5 MPa, mass flux of 200–2000 kg/m<sup>2</sup>s, inlet subcooling of 10–150 K. According to the experimental results, the location of ONB is identified based on the axial distribution of wall temperature, and the influence of various parameters on ONB in narrow rectangular channels is analyzed. It is found that heat flux, pressure, mass flux, and the gap size of the channel have a significant impact on ONB. By comparing the computed results of existing correlations, it is evident that there is a deviation, which can be attributed to the narrow range of experimental parameters in previous studies. Finally, a new ONB model is developed based on basic equations proposed by Hsu and the distribution of liquid temperature, taking into account the influence of mass flux and the enhanced heat transfer results from surrounding bubbles to correct the liquid temperature. The new correlation accurately describes the impact of each parameter and is in good agreement with the current experimental results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142315861","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-09-24DOI: 10.1016/j.anucene.2024.110914
Whether for shielding applications or for criticality safety studies, solving the neutron transport equation with good accuracy requires to take into account the resonant structure of cross sections in part of the Unresolved Resonance Region (URR). In this energy range even if the resonances can no longer be resolved experimentally, neglecting them can lead to significant numerical biases, namely in flux-based quantities. In Geant4, low energy neutrons are transported using evaluated nuclear data libraries handled by the Neutron High-Precision (Neutron-HP) package. In the version 11.01.p02 of the code, the URR can only be described by average smooth cross sections that do not take into account the statistical resonant structure of the cross sections. To overcome this shortcoming, the treatment of the URR with the use of the probability table method has been implemented in Geant4 and successfully validated with the reference Monte Carlo neutron transport codes MCNP6 (version 6.2) and Tripoli-4® (version 12). These developments will be taken into account in the next release of Geant4. All the validations of Geant4 have been performed with probability tables generated from both the NJOY and CALENDF pre-processing tools. Therefore Geant4 now has this unique feature to study the relative impact of the strategies involved during the production of probability table by the two pre-processing codes. This has been used to show that self-shielding is important also for inelastic cross sections in the example of 238U. The tool to generate probability tables usable by Geant4 either from NJOY or from CALENDF is made available on a dedicated GitLab repository and will be included in Geant4.
{"title":"Improvement of Geant4 Neutron-HP package: Unresolved resonance region description with probability tables","authors":"","doi":"10.1016/j.anucene.2024.110914","DOIUrl":"10.1016/j.anucene.2024.110914","url":null,"abstract":"<div><div>Whether for shielding applications or for criticality safety studies, solving the neutron transport equation with good accuracy requires to take into account the resonant structure of cross sections in part of the Unresolved Resonance Region (URR). In this energy range even if the resonances can no longer be resolved experimentally, neglecting them can lead to significant numerical biases, namely in flux-based quantities. In Geant4, low energy neutrons are transported using evaluated nuclear data libraries handled by the Neutron High-Precision (Neutron-HP) package. In the version 11.01.p02 of the code, the URR can only be described by average smooth cross sections that do not take into account the statistical resonant structure of the cross sections. To overcome this shortcoming, the treatment of the URR with the use of the probability table method has been implemented in Geant4 and successfully validated with the reference Monte Carlo neutron transport codes MCNP6 (version 6.2) and Tripoli-4® (version 12). These developments will be taken into account in the next release of Geant4. All the validations of Geant4 have been performed with probability tables generated from both the NJOY and CALENDF pre-processing tools. Therefore Geant4 now has this unique feature to study the relative impact of the strategies involved during the production of probability table by the two pre-processing codes. This has been used to show that self-shielding is important also for inelastic cross sections in the example of <sup>238</sup>U. The tool to generate probability tables usable by Geant4 either from NJOY or from CALENDF is made available on a dedicated GitLab repository and will be included in Geant4.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142315862","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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