Annals of Nuclear Energy最新文献

{"title":"Application of the PSI cycle check-up methodology for full-core Monte Carlo simulations verified using pin power estimates and validated for hot zero power conditions","authors":"Louis Berry ,&nbsp;Alexander Vasiliev ,&nbsp;Dimitri Rochman ,&nbsp;Mathieu Hursin ,&nbsp;Matthias Frankl ,&nbsp;Hakim Ferroukhi","doi":"10.1016/j.anucene.2024.110939","DOIUrl":"10.1016/j.anucene.2024.110939","url":null,"abstract":"<div><div>Monte Carlo (MC) based neutron transport codes provide high-fidelity numerical solutions to problems beyond the modelling capabilities of conventional core simulation methods. However, performing core-follow calculations with MC codes remains challenging due to the necessity of incorporating thermal-hydraulic feedback and generating evolved fuel composition for the cycle of interest by taking into account the entire irradiation history of loaded fuel assemblies over previous cycles. Nevertheless, at PSI, the neutronic version of a cycle check-up methodology (CHUP) is under development to address this challenge. This methodology involves extracting operating conditions and nuclide compositions from validated reference deterministic core-follow models (CASMO5/SIMULATE5/SNF), subsequently generating MC neutron transport models for selected operating points. This article presents the verification and validation performed for a hot zero power operating condition of a Swiss pressurised water reactor using the updated CHUP methodology, which automates information extraction from reference core models to minimise human intervention. The MC models were verified by assessing their deviation from criticality, consistently found to be supercritical within the [10,60] pcm range. Additionally, radial relative power distribution verification yield deviations in the [−5,4] % range compared to validated nodal results. Finally, additional validation was performed using start-up measurements from two successive cycles. Eight MC models were found to be subcritical, on average by 51 pcm, with a standard deviation of 25 pcm.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142357880","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}

{"title":"Neutron spectrum measurements inside a self-shielded cyclotron during 18F production using a single-cylindrical neutron spectrometer with gold and indium activation foils","authors":"Rasito Tursinah ,&nbsp;Sidik Permana ,&nbsp;Zaki Su’ud ,&nbsp;Alan Maulana ,&nbsp;A.K. Rivai ,&nbsp;Muhayatun Santoso ,&nbsp;Putu Sukmabuana ,&nbsp;Nurhuda ,&nbsp;Hari Suryanto ,&nbsp;Wahyudi ,&nbsp;Kusdiana ,&nbsp;Djoko Prakoso ,&nbsp;M.F. Ramadhani ,&nbsp;Bunawas","doi":"10.1016/j.anucene.2024.110949","DOIUrl":"10.1016/j.anucene.2024.110949","url":null,"abstract":"<div><div>The Single-Cylindrical Neutron Spectrometer (SCNS) is a single moderated spectrometer in the form of a cylinder with a diameter of 8 cm and a length of 16 cm. Seven Au activation foil type neutron detectors are installed on the moderator axis with varying distances. This spectrometer is used to measure the neutron spectrum in self-shielded cyclotron during ¹⁸F production using a proton of 11 MeV and a beam current of 25 µA for 70 min. The detector response was calculated using MCNPX 2.7. The measurement results obtained activity of ¹⁹⁸Au and ^116mIn and were validated with simulation using MCNPX. With the detector response values and detector activity from the measurements, a neutron spectrum was obtained using the unfolding technique using the UMG program. The neutron spectrum from measurements using SCNS with Au and In activation foil detectors has a difference of 8 %. Meanwhile, with the MCNPX calculation results, the measurement results using the Au foil are 1 % while with the In foil it is 10 %.</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":"142328130","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}

{"title":"Extending the finite elements neutronic code FENNECS to the Discontinuous Galerkin method","authors":"Romain Henry,&nbsp;Jérémy Bousquet,&nbsp;Armin Seubert","doi":"10.1016/j.anucene.2024.110927","DOIUrl":"10.1016/j.anucene.2024.110927","url":null,"abstract":"<div><div>The finite element method (FEM) neutronics code FENNECS was originally developed at GRS for unstructured geometry and first applied to fast reactors. It has recently been applied to light water reactors (LWRs). Indeed, most of the currently envisaged Small modular reactors (SMRs) concepts are based on well-proven LWR techniques. However, significant discrepancies were observed in terms of power distribution for the FENNECS prediction. The identified limitation is due to the implementation of the continuous Galerkin (CG) method, which enforces flux continuity between elements. This is incorrect due to the cross-section homogenization process. To overcome this problem, discontinuity factors can be introduced to allow for a more accurate description of the physics in LWR. Within the FEM frameworks, the modelling of discontinuous variables can be introduced using the discontinuous Galerkin (DG) formalism. FENNECS has been extended with the implementation of a new method based on the DG method. The method’s verification and validation were successfully achieved by comparing the assembly power distribution with a reference Monte Carlo solution for the NuScale SMR benchmark. Compared to the continuous approach, the discrepancies were significantly reduced from 20% to less than 4%.</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":"142328129","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}

{"title":"Layered target design method for global spectrum optimization of radioisotope production","authors":"Yu Xin ,&nbsp;Qingquan Pan ,&nbsp;Xiaojing Liu","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}

{"title":"Californium-252 production at the High Flux Isotope Reactor - II: Comparison between the highly enriched uranium and a proposed low-enriched uranium core","authors":"David Chandler,&nbsp;Donny Hartanto,&nbsp;Jin Whan Bae,&nbsp;Kevin M. Burg,&nbsp;Yves Robert,&nbsp;Carol Sizemore","doi":"10.1016/j.anucene.2024.110920","DOIUrl":"10.1016/j.anucene.2024.110920","url":null,"abstract":"<div><div>This is the second paper on a ²⁵²Cf production study performed in support of efforts to convert the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. The first paper primarily focuses on validating computational tools and nuclear data. This companion paper evaluates another critical aspect: the ²⁵²Cf production capability with a proposed LEU core. HFIR must maintain its world-class performance and missions following conversion and because ²⁵²Cf is a vital, multipurpose neutron-emitting radioisotope, the ability to efficiently produce ²⁵²Cf must be preserved. In this study, the HFIRCON transport and depletion tool, several nuclear data libraries, and Campaign 78 data were used to compute ²⁵²Cf production, sensitivity, and safety metrics. Results indicate the ²⁵²Cf production and production rates are slightly higher with a 95<!--> <!-->MW_th LEU core compared with those obtained with the 85<!--> <!-->MW_th HEU core. Additionally, the target peak fission rate densities, discharge cumulative fission densities, and heat deposition rates with the LEU core are within a few percent of those calculated with the HEU core. The findings suggest HFIR’s ²⁵²Cf production capability can be effectively maintained with an LEU core without adversely affecting the safety metrics.</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":"142328127","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}

{"title":"Investigation on cross-flow characteristics of two different mixing vane grid in a 5 × 5 rod bundle","authors":"Li Jiang,&nbsp;Jianqiang Shan,&nbsp;Shan Zhou,&nbsp;Yixiang Zou,&nbsp;Junliang Guo,&nbsp;Miao Gui","doi":"10.1016/j.anucene.2024.110934","DOIUrl":"10.1016/j.anucene.2024.110934","url":null,"abstract":"<div><div>To comprehend the combined mixing effects downstream two different MVGs installed at same axial height, individual analysis of these MVGs is the precondition. In this paper, an experimental study on the crossflow distribution downstream two different split-type MVGs is presented by PIV in a 5 × 5 rod bundle. The cross flow in the sixteen inner subchannels is obtained at elevation from 1.8 <em>D_h</em> to 25 <em>D_h</em> downstream grids. For both two types of grids, two individual co-rotating vortices are observed near the grid in each subchannel and inter-subchannel cross flow increases before approximate 3 <em>D_h</em> and decays exponentially. The secondary-flow and cross-flow intensity of grid type Ⅰ overweigh the type Ⅱ and the intensities are corelated to horizontal projected area of vane. Based on the experimental results, an improved subchannel mixing grid model is proposed and the predicted cross flow in the gap is in good agreement with measured results.</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":"142328128","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}

{"title":"Assessing organizational vulnerability of nuclear power plants using AHP-fuzzy sets method","authors":"Ye Wang ,&nbsp;Huiwei Wei ,&nbsp;Jing Wen ,&nbsp;Jiayuan He ,&nbsp;Pengcheng Li","doi":"10.1016/j.anucene.2024.110896","DOIUrl":"10.1016/j.anucene.2024.110896","url":null,"abstract":"<div><div>The operational efficiency and developmental progress of nuclear power entities are significantly challenged by organizational vulnerability, which can lead to severe outcomes if neglected. This paper presents a systematic approach to identifying and quantifying the primary factors influencing organizational vulnerability within nuclear power plants (NPPs). An evaluative index is established, and an innovative hybrid methodology combining the Analytic Hierarchy Process (AHP) and fuzzy sets theory is applied to assess overall organizational vulnerability. A case study validates the approach, demonstrating low vulnerability and strong organizational reliability in the NPPs. The research contributes a valuable tool for enhancing safety and sustainability in the nuclear power sector.</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":"142328126","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}

{"title":"Griffin: A MOOSE-based reactor physics application for multiphysics simulation of advanced nuclear reactors","authors":"Yaqi Wang ,&nbsp;Zachary M. Prince ,&nbsp;Hansol Park ,&nbsp;Olin W. Calvin ,&nbsp;Namjae Choi ,&nbsp;Yeon Sang Jung ,&nbsp;Sebastian Schunert ,&nbsp;Shikhar Kumar ,&nbsp;Joshua T. Hanophy ,&nbsp;Vincent M. Labouré ,&nbsp;Changho Lee ,&nbsp;Javier Ortensi ,&nbsp;Logan H. Harbour ,&nbsp;Jackson R. Harter","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}

{"title":"Steady-state thermal–hydraulic analysis of an NTP reactor core based on the porous medium approach","authors":"Zichao Han,&nbsp;Jing Zhang,&nbsp;Mingjun Wang,&nbsp;Wenxi Tian,&nbsp;Guanghui Su,&nbsp;Suizheng Qiu","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}

{"title":"Experimental study on the plate-type fuel melting behavior based on alternative materials","authors":"Zhiyuan Wu,&nbsp;Kepiao Li,&nbsp;Kui Zhang,&nbsp;Ronghua Chen,&nbsp;Wenxi Tian,&nbsp;Suizheng Qiu","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}