Nuclear Engineering and Design最新文献

{"title":"Spent fuel pool boil-off experiments with high power and high time resolution","authors":"Christoph Schuster ,&nbsp;Oleg E. Stepanov","doi":"10.1016/j.nucengdes.2025.113958","DOIUrl":"10.1016/j.nucengdes.2025.113958","url":null,"abstract":"<div><div>The occurrence of spent fuel pool (SFP) accidents, in which the water level decreases below the top of the fuel assemblies (FA), may arise due to the loss of active pool cooling or pool leakage. In such scenarios, the ALADIN test facility of the TUD Dresden University of Technology is utilized to simulate a comprehensive boiling water reactor (BWR) FA in a 1:1 scale. This report supplements the existing data base with rod powers between 100 W and 350 W. Boil-off experiments will be analyzed in detail in relation to axial and radial temperature profiles as a function of time and heater power up to a maximum temperature of approximately 800 °C. Additionally, temperature excursion mitigating measures such as spray-cooling and flooding will be discussed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113958"},"PeriodicalIF":1.9,"publicationDate":"2025-03-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143552853","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":"Investigation of decay heat removal systems in the ALLEGRO helium-cooled fast reactor","authors":"Zsombor Bali ,&nbsp;Gusztáv Mayer","doi":"10.1016/j.nucengdes.2025.113952","DOIUrl":"10.1016/j.nucengdes.2025.113952","url":null,"abstract":"<div><div>ALLEGRO is a demonstrator reactor of the GFR2400 helium-cooled gas fast reactor selected for future research by the Generation IV International Forum (GIF) in the early 2000s. This 75 MW thermal power demonstrator reactor has been developed in several European projects over the past two decades. The current ALLEGRO design consists of pressurized water on the secondary side of both the main heat exchangers (MHX) and the decay heat removal systems (DHR). Nevertheless, in case of a primary-secondary break at the MHXs or the DHRs, the water coolant may ingress into the gas-cooled reactor core which may cause reactivity and corrosion issues. For this reason, replacing the secondary water in the MHXs and the DHRs is an important topic during the design phase of ALLEGRO. In this paper, the gases of helium, nitrogen and CO₂ are investigated as possible secondary coolants of the DHRs at different pressures using the CATHARE thermal hydraulics system code. The results show that from a cooling performance point of view, CO₂ gas is a promising candidate and by using this gas, a relatively small design change is needed concerning the size of the DHR heat exchangers.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113952"},"PeriodicalIF":1.9,"publicationDate":"2025-03-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534355","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":"Steam generator model design parameter sensitivity study for small modular reactor system","authors":"Palash K. Bhowmik ,&nbsp;Congjian Wang ,&nbsp;Nicholas Hernandez ,&nbsp;Tejas Kedlaya ,&nbsp;Piyush Sabharwall","doi":"10.1016/j.nucengdes.2025.113973","DOIUrl":"10.1016/j.nucengdes.2025.113973","url":null,"abstract":"<div><div>This study focuses on design parameter sensitivity studies pertaining to several Once-Through Steam Generator (OTSG) model cases both with and without a riser using python and advanced risk assessment and optimization tool, i.e. Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory (INL), to support a Small Modular Reactor (SMR) system. The presented Steam Generator (SG) python-based model is a mathematical representation of a steam-generating unit for a Pressurized Water Reactor (PWR)-type SMR system, including fluid flow and heat transfer equations, models, and correlations. Design studies involve changing the model’s input design parameters (e.g., temperature, pressure, mass flow rate) to observe the resulting effects on the output of the system, such as the Heat Transfer Coefficient (HTC), Reynolds number, Nusselt number, and heat transfer performance. Sensitivity studies analyze the degree to which system output and/or desired parameters (e.g., HTC or heat transfer performance) are sensitive to changes in the input parameters. By using RAVEN, detailed design parametric sensitivity studies. Six input parameters—namely, the pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid) of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (i.e., ±1%, ±5% and ±10 % relative changes) for 600 samples. The analysis results give valuable insights into SG system performance, and provide justification for further research and development such as optimized sensor placement, design verification, validation, and optimization.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113973"},"PeriodicalIF":1.9,"publicationDate":"2025-03-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143534354","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":"Neutronic uncertainty propagation for the SAFARI-1 research reactor benchmark: IAEA CRP T12029","authors":"O.E. Montwedi ,&nbsp;V.V. Naicker ,&nbsp;G.P. Nyalunga","doi":"10.1016/j.nucengdes.2025.113943","DOIUrl":"10.1016/j.nucengdes.2025.113943","url":null,"abstract":"<div><div>Quantification of full core neutronic and sensitivity analysis of the SAFARI-1 research reactor was performed; the output parameter of interest was the neutron multiplication factor k_eff. Fuel element calculations were also studied to gain insights into the problem. The calculational approach in the Organization for Economic Corporation and Development benchmark for quantification of uncertainties in multiphysics calculations for LWRs was followed. The scope of the work is limited to reactor physics (neutronic calculations). Five core models were built in line with the SAFARI-1 benchmark specifications, of the 2016 IAEA Coordinated Research Programme. Model 1 was a Fuel Element (FE) model with reflective boundary conditions on the sides and void boundary conditions axially. Models 2 – 4 were full core models with control rods at 0<!--> <!-->%, 50<!--> <!-->% and 100<!--> <!-->% insertion in the core. Model 5 was a full core model consisting of only fuel elements.</div><div>TSUNAMI-3D, which is part of the SCALE 6.2.3 code package, was used for uncertainty propagation. The uncertainty in the multiplication factor k_eff is calculated with the sensitivity coefficients and the covariance matrix using the so-called “sandwich rule”. It can be seen from the results that the uncertainty in k_eff is 0.490<!--> <!-->% for the infinite fuel element model and about 0.640<!--> <!-->% (highest value) for the full core model. It is also seen from the sensitivity analysis that the highest contributors to the uncertainty in the FE model are due to the average number of neutrons released per fission and due to the (n, gamma) reactions. For the full core model, the highest contributors are uncertainties due to the average number of neutrons per fission and that due to the fission spectrum, chi. A detailed behaviour of the individual contributors to the uncertainty is also investigated. It was also found that the size of the reactor core influences the value of the individual contributors. To a lesser extent, the heterogeneity of the reactor core also influences the value of the individual contributors. The neutron flux distribution in cores of different sizes was also studied since it contributes to the uncertainty due to chi, as well as the neutron importance, to observe the trend in the individual uncertainty contributions of chi and nubar.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113943"},"PeriodicalIF":1.9,"publicationDate":"2025-03-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143529732","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":"Study on high-resolution activation analysis based on the Monte Carlo global variance reduction method","authors":"Leiming Li ,&nbsp;Zhenping Chen ,&nbsp;Aikou Sun ,&nbsp;Chao Yang ,&nbsp;Tao Yu","doi":"10.1016/j.nucengdes.2025.113961","DOIUrl":"10.1016/j.nucengdes.2025.113961","url":null,"abstract":"<div><div>During nuclear reactor operations, neutron activation reactions generate significant quantities of radionuclides from the structural materials, directly impacting shielding design, maintenance planning, and decommissioning strategies. This is a critical component of radiation safety analysis. As advanced nuclear reactor technology evolves, the increasing complexity of reactor geometry, material configurations, and neutron spectra complicates activation analysis. Consequently, there is a pressing need for high-resolution activation analysis of nuclear reactors. This paper presents a high-resolution activation analysis method for large-scale complex structural materials, utilizing the Monte Carlo global variance reduction particle transport technique. A fully automated coupled high-resolution activation analysis program is developed, enabling the calculation of high-resolution decay source distributions and precise evaluation of decay photon sources for extensive complex structural materials. The methodology is benchmarked against the shutdown dose rate benchmark released by the International Thermonuclear Experimental Reactor (ITER) program. Additionally, an application study of high-resolution activation analysis is conducted on a standard pressurized water reactor (PWR). The methodology demonstrated in this paper holds significant engineering value, enhancing the accuracy of activation calculations for large-scale nuclear reactor structural materials and it provides guidance for optimizing shielding design to reduce radiation exposure during operation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113961"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520204","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":"Enhancing uncertainty analysis: POD-DNNs for reduced order modeling of neutronic transient behavior","authors":"Yijun Zhang ,&nbsp;Wenhuai Li ,&nbsp;Sitao Peng ,&nbsp;Jinggang Li ,&nbsp;Ting Wang ,&nbsp;Qingyun He ,&nbsp;Tao Wang ,&nbsp;Haoliang Lu ,&nbsp;Ling Zeng","doi":"10.1016/j.nucengdes.2025.113969","DOIUrl":"10.1016/j.nucengdes.2025.113969","url":null,"abstract":"<div><div>In reactor safety analysis, sensitivity analyses on critical parameters are essential for ensuring the reliability of safety conclusions, particularly regarding transient behavior, which often requires time-consuming computations. Developing surrogate models presents a promising solution. This paper extends the Proper Orthogonal Decomposition-Radial Basis Function (POD-RBF) framework to the 3D Light Water Reactor Core Transient Benchmark (3DLWRCT) for control rod ejection accidents. The primary aim is to simulate transient behavior under random perturbations in the macroscopic neutronic cross-sections of fuel assemblies.</div><div>Our results indicate that the traditional POD-RBF approach struggles to accurately reconstruct the highly nonlinear transient system, whether through spatiotemporal folding or spatial data reduction. To overcome these challenges, we enhance the model by integrating Deep Neural Networks (DNNs) and employing the Tree-structured Parzen Estimator for optimal neural network architecture selection. This improved approach significantly increases the accuracy of the surrogate models, demonstrating its feasibility and effectiveness. The integration of DNNs offers a deeper understanding of complex interactions within the reactor core, effectively capturing nonlinearities and yielding reliable predictions even under uncertainty.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113969"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520205","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":"VVER long-term operation – A review based on the material studies results from past and ongoing EU-supported research projects","authors":"Vladimir Slugen ,&nbsp;Jana Simeg Veternikova ,&nbsp;Maria Domankova ,&nbsp;Matus Gavalec ,&nbsp;Jana Petzova ,&nbsp;David Slnek ,&nbsp;Mykola Dzubinsky ,&nbsp;Oleksii Shugaylo ,&nbsp;Ildiko Szenthe ,&nbsp;Ferenc Gillemot ,&nbsp;Maksym Zarazovskii ,&nbsp;Jari Lydman ,&nbsp;Radim Kopriva ,&nbsp;Szabolcs Szavai ,&nbsp;Ondrej Srba","doi":"10.1016/j.nucengdes.2025.113949","DOIUrl":"10.1016/j.nucengdes.2025.113949","url":null,"abstract":"<div><div>Safe and long-term operation (LTO) of nuclear power plants (NPP) with Water-Water Energetic Reactors (VVER) is essential for several central and eastern European countries to keep their energy supply security. During the last decades, several EU-supported projects in framework projects or Horizon2020 schemes have focused on the material degradation of pressurised water reactors (PWRs), especially VVERs. This paper aims to extract the most important results dominantly from EURATOM projects which support or could limit the long-term operation of this Soviet design of PWRs. The accent is given on ongoing projects such as FRACTESUS, STRUMAT-LTO, APAL, and ENTENTE. Actually, running project DELISA-LTO (2022–2026) comprises not only results from previous research in this area but also incorporates wide material databases coming from nuclear power plant V-1 in Jaslovské Bohunice (Slovakia), which was shut down after 28 years of operation and provides wide-range of original service-aged materials specimens for profound material studies having in mind lifetime extension of VVERs. Consolidated knowledge focused mainly on the reactor pressure vessel (RPV) material studies is summarised and supported by relevant references.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113949"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143526997","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":"[NURETH-20] Evaluation on mitigation performance of flooding safety system under hypothetical loss of coolant accident in Korean i-SMR with MELCOR code","authors":"Chang Hyun Song ,&nbsp;Jae Hyung Park ,&nbsp;JinHo Song ,&nbsp;Sung Joong Kim","doi":"10.1016/j.nucengdes.2025.113950","DOIUrl":"10.1016/j.nucengdes.2025.113950","url":null,"abstract":"<div><div>In response to the growing interest in Small Modular Reactors (SMRs) globally, many countries are actively pursuing the development of high-power SMRs, based on their inherent advantages including enhanced safety, grid flexibility, and potential for hydrogen production. Among these endeavors, an i-SMR with an electrical power output of 170 MWe has been under development since 2021 by Korea Hydro &amp; Nuclear Power Co., Ltd. in Republic of Korea. The i-SMR has established ambitious top-tier requirements, such as core damage frequency less than 1.0 × 10⁻⁹/module-year and large early release frequency less than 1.0 × 10⁻¹⁰/module-year, and emergency planning zone within nuclear power plant site boundary. All of which is extremely challenging and necessitates innovative safety systems with exceptional reliability. In this context, this study proposed a Flooding Safety System (FSS) as a novel safety system, and its mitigating performance under a hypothetical accident scenario was evaluated by using MELCOR code for validating the efficacy of this conceptual approach. To conduct the accident analysis, a MELCOR input model for the i-SMR was developed. The initiating event assumed was the stuck open of depressurization valve, leading to the discharge of coolant from the primary system into the metal containment vessel, which can be deemed as a loss of coolant accident. The findings in this study revealed that timely operation of the FSS can prevent core damage, thus validating its crucial role to assuring the integrity and reliability of the i-SMR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113950"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143510666","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":"Impact of hydrogen on iPWR containment Thermal-Hydraulics in severe accidents","authors":"Luke Lebel,&nbsp;Dening Eric Jia,&nbsp;David Grand-Maitre,&nbsp;Tony Clouthier","doi":"10.1016/j.nucengdes.2025.113953","DOIUrl":"10.1016/j.nucengdes.2025.113953","url":null,"abstract":"<div><div>Addressing hydrogen issues is an important part of severe accident management for nuclear accidents, and this is doubly important for water-cooled small modular reactors (SMRs), like integral pressurized water reactors (iPWRs). This study uses a combination of containment thermal–hydraulic experiments and modeling to assess the impact of introducing light non-condensable gases under prototypic accident conditions. The experiments evaluated the gas mixing behaviour and steam condensation heat transfer coefficients in the tertiary steam–air-helium system, extending the unified steam condensation correlations by <span><span>Dehbi (2016)</span></span> past the density difference transition point. The modeling with the GOTHIC thermal-hydraulics code scaled the experimental observations to reactor-relevant accident scenarios, with steam-hydrogen in a closed containment and steam–air-hydrogen in a loss of containment isolation situation. A strong reduction in the heat transfer capacity caused by the presence of non-condensable gases was observed, which disrupt the steam condensation that drives many of the passive cooling loops that form the basis of accident mitigation strategies. Likewise, the observed gas mixing behaviour inferred that hydrogen combustion may be a further issue that needs to be addressed in severe accident management guidelines.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113953"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143510665","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":"Prediction of radiation shielding design schemes based on adaptive neural networks","authors":"Qisheng Chen ,&nbsp;Zi-Hui Yang ,&nbsp;Zhong-Yang Li ,&nbsp;Guo-Min Sun ,&nbsp;Shi-Peng Wang ,&nbsp;Yu-Chen Li ,&nbsp;Zhi-Xing Gu ,&nbsp;Fei Li ,&nbsp;Juan Fu ,&nbsp;Gui-Hua Tao","doi":"10.1016/j.nucengdes.2025.113933","DOIUrl":"10.1016/j.nucengdes.2025.113933","url":null,"abstract":"<div><div>As space exploration and space reactor technology continue to advance, radiation shielding design faces numerous challenges, such as space limitations, weight constraints, and the complexity of shielding materials. Traditional design methods typically rely on empirical models validated through Monte Carlo simulations, but these approaches often fail to achieve optimal results. To enhance the radiation protection efficiency of space reactors, this paper proposes a deep neural network model based on the self-attention mechanism to assist in predicting radiation shielding design schemes and to verify its accuracy and practicality. We used SuperMC software to record the relevant safety parameters for the Kilopower reactor under 10,000 different shielding design schemes and calculated the total mass and total radiation dose for these designs, creating a comprehensive dataset. The total mass and radiation dose were used as inputs to the neural network, which then generated the corresponding radiation shielding design schemes. Experimental results show that the model demonstrates high accuracy and strong interference resistance, with the error in total radiation dose and material mass consistently controlled around 3%. Additionally, by combining simulation methods with the self-attention mechanism, the model effectively generates radiation shielding designs suitable for space reactors, providing reliable protection solutions for future space missions. This approach also opens new possibilities for radiation shielding design in other fields.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113933"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143511332","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}