Nuclear Engineering and Design最新文献

{"title":"Measurement of temperature distribution and estimation of interfacial heat flux profiles during the interaction of different molten metals with ferrosiliceous concrete","authors":"Pedduri Jayakrishna ,&nbsp;Akshay U. Shirsat ,&nbsp;Prakash Nanthagopalan ,&nbsp;Arunkumar Sridharan ,&nbsp;Shyamprasad Karagadde ,&nbsp;Anuj Kumar Deo ,&nbsp;Srinivasa Rao ,&nbsp;P.K. Baburajan ,&nbsp;S.V. Prabhu","doi":"10.1016/j.nucengdes.2025.114099","DOIUrl":"10.1016/j.nucengdes.2025.114099","url":null,"abstract":"<div><div>The necessity of understanding the interaction of molten corium with the sacrificial concrete layer surrounding the nuclear reactors has been the motivation to perform MCCI experiments. Materials (zinc, aluminium and stainless steel) which have different melting temperatures and thermal properties are chosen and the experimental investigations are carried out by pouring molten metals in well-defined cavities of concrete test sections to study the ablation and thermal behaviour of the concrete. Coarse and fine aggregates of hematite are added to the cement to attain the required mechanical and thermal properties and provide better radiation shielding. The transient interaction of the molten metal with the concrete is measured using thermocouples. The measured subsurface temperatures are subsequently used to estimate the interfacial heat flux profiles by solving an inverse heat conduction problem using the sequential function specification method. The peak temperatures measured and peak heat flux values estimated in concrete samples interacted with zinc, aluminium and stainless steel are around 85 °C and 35 <span><math><mrow><mi>k</mi><mi>W</mi><mo>/</mo><msup><mrow><mi>m</mi></mrow><mn>2</mn></msup></mrow></math></span>, 105 °C and 58 <span><math><mrow><mi>k</mi><mi>W</mi><mo>/</mo><msup><mrow><mi>m</mi></mrow><mn>2</mn></msup></mrow></math></span>, and 468 °C and 65 <span><math><mrow><mi>k</mi><mi>W</mi><mo>/</mo><msup><mrow><mi>m</mi></mrow><mn>2</mn></msup></mrow></math></span>, respectively. There is no ablation during the interaction of zinc and aluminium with the concrete test sections. Though the melting temperature of stainless steel (1450 °C) is higher than the ablation temperature of the concrete (around 1200 °C), concrete did not undergo significant ablation due to the higher density and the strength offered by the addition of hematite aggregates. The results presented in the current study are accurate within the time period where the semi-infinite model is valid.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114099"},"PeriodicalIF":1.9,"publicationDate":"2025-05-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143898917","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":"Evaluation of radiation shielding properties of concrete with oil shale ash and basalt-boron fiber additives for spent nuclear fuel casks","authors":"Andrius Slavickas ,&nbsp;Hando Tohver ,&nbsp;Maryna Holiuk ,&nbsp;Andrejs Krasnikovs ,&nbsp;Riho Mõtlep ,&nbsp;Volodymyr Gulik","doi":"10.1016/j.nucengdes.2025.114110","DOIUrl":"10.1016/j.nucengdes.2025.114110","url":null,"abstract":"<div><div>Cement-based concretes are commonly used for managing radioactive waste due to their low cost and acceptable radiation shielding properties. However, the potential of using oil shale ash as an admixture in concrete for the radioactive waste packaging has been left unnoticed. The present work aims to demonstrate the radiation shielding properties of developed concrete mixes with the oil shale ash and basalt-boron fibers for the spent nuclear fuel management at the Ignalina NPP. Concrete composite materials were placed in the CONSTOR RBMK-1500/M2 cask model to estimate radiation shielding properties. OpenMC, MCNP, SCALE and Serpent codes were used for the estimation of radiation shielding properties. It was demonstrated that the cask with developed concrete composite material can meet the safety limit.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114110"},"PeriodicalIF":1.9,"publicationDate":"2025-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143890522","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":"The thorny question of critical heat flux in plate-type reactors: A coherent methodology for the PALLAS-reactor","authors":"F. Bertocchi ,&nbsp;C.A. Vega","doi":"10.1016/j.nucengdes.2025.114075","DOIUrl":"10.1016/j.nucengdes.2025.114075","url":null,"abstract":"<div><div>The PALLAS-reactor is an advanced, plate-type nuclear reactor designed for producing medical isotopes currently under construction in the Netherlands. Adequately estimating the critical heat flux (CHF) for all postulated accident scenarios is crucial for licensing the reactor, for which the Reactor Excursion and Leak Analysis Program (RELAP5/MOD3.3 Patch 5) program is the primary licensing code. Literature has shown that the look-up tables (LUT) of RELAP over-predict the CHF with forced convection through narrow rectangular channel, thus highlighting the need for alternative tools better suited to this geometry. These tools have been identified in ancillary publications complementary to the present one. However, a methodology is still missing that provides guidance to the analyst for estimating the CHF in all postulated accident conditions. Therefore, this publication aims at providing a first coherent approach to estimate the CHF in the rectangular channels of a plate-type nuclear reactor, by illustrating the case study of the PALLAS-reactor. Of the available CHF correlations, we have selected the Sudo-Kaminaga correlation modified by Kim (SK-Kim) for estimating the CHF in forced convection. For natural circulation, the CHF is estimated through the 2006 LUT of RELAP5/MOD3.3 Patch 5 since these are applicable to narrow channels when buoyancy is dominant. Discriminating between forced and natural convection regimes relies on the value of a dimensionless group derived from representative experiments. If the SK-Kim correlation predicts that the CHF limit is approached in one of the cooling channels, this is to be modeled by a validated computational fluid dynamics (CFD) tool in order to more accurately account for the effects of fast transients. This work provides the first, well-founded and coherent methodology for estimating the CHF in a plate-type nuclear reactor. In this respect, it represents a cornerstone of the PALLAS-reactor licensing process that, just as importantly, can support the safety analyses of other reactors of similar design.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114075"},"PeriodicalIF":1.9,"publicationDate":"2025-04-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143886740","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":"Local and global sensitivity analyses on the overflow probability of Abadia de Goiás radioactive near surface repository","authors":"D.G. Teixeira ,&nbsp;P.F. Frutuoso e Melo ,&nbsp;L.G.M. Alvim ,&nbsp;A.C.M. Alvim ,&nbsp;A.S.M. Alves","doi":"10.1016/j.nucengdes.2025.114090","DOIUrl":"10.1016/j.nucengdes.2025.114090","url":null,"abstract":"<div><div>This work presents a sensitivity analysis of the overflow probability of the near- surface radioactive repository built after the Abadia de Goiás accident, where radioactive ¹³⁷Cs was released. The purpose of the analysis is to help the repository owner and the regulatory body make decisions on design parameters that affect the overflow probability. The following parameters have been considered: degradation function of the repository ceiling, internal porosity of the repository, internal area of the repository base, repository base and wall thicknesses, repository height, height of the liquid column inside the repository, average rainfall rate, evapotranspiration rate, surface runoff and hydraulic conductivity of the repository concrete. The choice for the range of each parameter is discussed. An investigation of the role of the mentioned parameters is detailed based on results obtained in a previous paper, and the dependencies among the rate parameters on the average rainfall rate is emphasized and modeled. To give a deeper perspective, local and global sensitivity analyses have been performed. In the case of global analysis, two approaches have been followed: in the first, the concept of Normalized Mutual Information and the other one leverages Shapley values derived from Light Gradient Boosting Machine and Shapley values have been used for investigating non-linearities. For some parameters the lowest values are to be considered, e.g., the repository base thickness and the average rainfall rate, while for others the highest value is desirable, like the evapotranspiration rate and the surface runoff in order to obtain lower overflow probabilities. These results should be looked with care because nonlinearities have been identified by the Shapley value analysis. Overall, the degradation function and the evapotranspiration rate emerge as the most influential parameters. The legacy of this work lies in the robust methodological framework it provides, paving the way for enhanced safety assessments and improved design practices in complex environmental systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114090"},"PeriodicalIF":1.9,"publicationDate":"2025-04-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143891279","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":"Burnup calibration and its dependence on irradiation history in high-temperature gas-cooled reactors","authors":"Hongjian Zhang,&nbsp;Liguo Zhang,&nbsp;Yu Wang,&nbsp;Yanlong Wen,&nbsp;Qing Zhu,&nbsp;Tao Ma","doi":"10.1016/j.nucengdes.2025.114078","DOIUrl":"10.1016/j.nucengdes.2025.114078","url":null,"abstract":"<div><div>Accurate online burnup measurement in pebble-bed high-temperature gas-cooled reactors (HTGRs) is a critical technical prerequisite for continuous refueling. This study provided a theoretical analysis of the online burnup measurement methodology for the HTR-PM (High-Temperature Gas-Cooled Reactor Pebble-Bed Module), focusing on the calibration between burnup and Cs-137 activity. Specifically: (1) A general burnup calibration is derived that correlates burnup with Cs-137 activity, accounting for contributions from multiple fissile isotopes. The approximated burnup obtained using this burnup calibration exhibit negligible deviations from the theoretical values. (2) Potential sources of error, such as differences in the fission energy and fission yields of various fissile isotopes, are analyzed.</div><div>To evaluate the applicability of the general burnup calibration and quantify the effects of irradiation history, a multi-cycle frequently-varying irradiation history model was developed. Leveraging KORIGEN and Nuclear Inventory Tool (NUIT) as computational tools, extensive burnup data were generated to establish a fitted burnup calibration. The study then: (1) Compared the general burnup calibration with the fitted burnup calibration. (2) Quantified the impact of frequently-varying irradiation history on burnup calibration accuracy.</div><div>Burnup can be conveniently estimated using a linear correlation with Cs-137 activity. This study builds upon that foundation by introducing theoretical and computational refinements, improved accuracy under long irradiation periods. The results reveal that the relative error between the NUIT-fitted burnup calibration and the general burnup calibration is less than 0.55 %. Additionally, the effect of irradiation history variations on burnup calibration is so small that under identical fuel burnup conditions, the standard deviation of Cs-137 activity is less than 4.0 × 10⁸ Bq, with a coefficient of relative variation not exceeding 0.60 %. These findings provided valuable insights for accurately predicting fuel sphere burnup behavior, optimizing fuel management strategies, and ensuring the safe and efficient operation of HTGRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114078"},"PeriodicalIF":1.9,"publicationDate":"2025-04-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143886739","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":"Development of a transient analysis model for liquid metal cooled space nuclear reactor power system","authors":"Li Huaqi,&nbsp;Jiang Xinbiao,&nbsp;Tian Xiaoyan,&nbsp;Zhu Lei,&nbsp;Shi Leitai,&nbsp;Chen Sen,&nbsp;Luo Xiaofei,&nbsp;Li Da,&nbsp;Chen Lixin","doi":"10.1016/j.nucengdes.2025.114089","DOIUrl":"10.1016/j.nucengdes.2025.114089","url":null,"abstract":"<div><div>The space nuclear reactor power system (SNRPS), which uses a liquid metal-cooled reactor coupled with a closed-loop Brayton cycle for thermoelectric conversion, has emerged as the preferred choice for an advanced space power system. This preference is due to its excellent heat transfer efficiency in the core, wide power range, controllable equipment size, high power conversion efficiency, and mature technology. This study focuses on establishing a system transient analysis model for reactor design, control, and safety analysis of a SNRPS. The model includes subsystem models such as thermal–hydraulic models for the reactor core and primary coolant loop systems, power conversion unit models, and heat pipe radiator models. Six types of benchmark test problems are used to validate each sub-module and component model. The results show that the maximum absolute error between the sub-module model and the analytical solutions of these benchmark test problems is within 2%. Based on the theoretical model established, a transient analysis code for the space nuclear reactor (TACSNR) was developed. The TACSNR was verified using steady-state design parameters from the ultra-small liquid metal cooled space nuclear reactor power system concept (ULCR SNRPS) and the inherent safety sectored compact reactor with a SiGe thermoelectric (TE) power conversion assembly space nuclear reactor power system (SCoRe-TE SNRPS) startup transient process. The calculation results show that the maximum absolute deviation between the calculated values of the TACSNR and the steady-state design parameters of the ULCR SNRPS conceptual scheme is less than 1%, consistent with the parameter change trend and numerical values during the transient startup process of the SCoRe-TE SNRPS system. Additionally, the maximum relative deviation at rated steady state of the SCoRe-TE SNRPS is less than 12%.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114089"},"PeriodicalIF":1.9,"publicationDate":"2025-04-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143890521","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":"Development of point model for recombiner and its validation with tests conducted in THAI facility using CFD","authors":"Rupak Kumar Raman ,&nbsp;R.S. Rao ,&nbsp;Kannan N. Iyer ,&nbsp;Sanjeev Gupta","doi":"10.1016/j.nucengdes.2025.114100","DOIUrl":"10.1016/j.nucengdes.2025.114100","url":null,"abstract":"<div><div>A point model for a recombiner compatible for CFD code to simulate hydrogen distribution and mitigation in nuclear power plant containments is developed in this study. Given the significantly smaller size of the recombiner compared to containment compartments, fully resolving the recombination process requires detailed geometric modeling and a highly refined mesh within the recombiner channels, which require high computational demand. Present study simplifies the recombiner channel into a single computational node while ensuring conservation of momentum, energy and species to address the issue of computational demand. A set of correlations for pressure drop, heat generation and heat transfer coefficients has been developed for recombiner channel in the presence of a steam environment through parametric studies. This model is incorporated in CFD code FLUENT using UDF. Through various simulations the derived point model is verified under both steady-state and transient condition with the detailed computational model. Further validation is performed using CFD simulations of the HR-49, HR 2 and HR 5 tests from the THAI experiment program for various test conditions. The simulation with the point model successfully predicted the recombiner performance and species concentration with experimental data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114100"},"PeriodicalIF":1.9,"publicationDate":"2025-04-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143882779","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":"Loss-of-flow transients in the liquid metal-cooled reactor-pool experiment E-SCAPE","authors":"Katrien Van Tichelen ,&nbsp;Fabio Mirelli ,&nbsp;Yann Bartosiewicz ,&nbsp;William D’haeseleer","doi":"10.1016/j.nucengdes.2025.114086","DOIUrl":"10.1016/j.nucengdes.2025.114086","url":null,"abstract":"<div><div>Natural circulation in the primary system is one of the key mechanisms for removing the decay heat in the MYRRHA pool-type research reactor under development at SCK CEN, the Belgian Nuclear Research Centre. To confirm the feasibility of this passive approach, experiments are performed in the E-SCAPE facility, a thermal hydraulic 1/6-scale 3-D model of the primary system of MYRRHA, with an electrical core simulator and cooled with Lead Bismuth Eutectic.</div><div>This paper presents the outcome of transient loss-of-flow (LOF) experiments in E-SCAPE. First, the representativeness of LOF transients in E-SCAPE for MYRRHA is demonstrated based on simplified analytical integral models of the reactor prototype and of the scaled facility. Next, results of several test cases with varying core powers and system pressure losses are reported. In all cases studied, a smooth transition from forced to buoyancy-driven natural circulation is observed after the LOF event, with the establishment of stable, lower flow rates. Decay heat can be safely removed from the core as the maximum core temperatures stay within safety limits. Two phases can be identified during the transient: an initial phase dominated by mass inertia, and a second phase dominated by the heat capacity of the system. The final steady state shows significant thermal stratification in the upper plenum.</div><div>The extensive instrumentation in the E-SCAPE facility allows direct comparison of experimental data with numerical simulations, allowing validation of simulation tools in representative conditions. This is essential for the safety assessment and licensing process of MYRRHA.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114086"},"PeriodicalIF":1.9,"publicationDate":"2025-04-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143882100","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":"Simulating experimentally observed nonlinear response of large-scale concrete structure to understand the selection of damping: A case of minor nonlinearities","authors":"Sangwoo Lee,&nbsp;Abhinav Gupta,&nbsp;Giorgio T. Proestos","doi":"10.1016/j.nucengdes.2025.114098","DOIUrl":"10.1016/j.nucengdes.2025.114098","url":null,"abstract":"<div><div>Recent studies conducted by the US Nuclear Regulatory Commission and its collaborators have explored the use of limit-state C for SDCs 5 and 4, unlike the conventional design of concrete shear walls in nuclear power plants. Consideration of the limit-state C allows minor nonlinearity in the behavior of structural systems when subjected to design earthquakes. In the context of the nonlinear behavior in concrete structures, the selection of appropriate parameters for the concrete’s constitutive material model is important. In addition, there are some concerns with using Rayleigh damping in nonlinear seismic analysis because, many studies have shown that an improper use of Rayleigh damping in the nonlinear seismic analysis can lead to unintended large damping forces thereby resulting in an underestimation of response parameters. In this study, response data from a large-scale shake table experiment of a 3-story concrete shear wall structure is used to understand these effects. A finite element analysis of the test specimen using concrete damage plasticity model and its reconciliation with the experimental data is used to understand two aspects discussed above, i.e., (i) selection of model parameters in the Concrete Damage Plasticity Model for nonlinear seismic analysis of concrete structures, and (ii) selection of an appropriate damping model. Both of these aspects are studied for the case of minor damage (nonlinearity) in the structure corresponding to ASCE-43′s guidelines for risk-informed performance-based design.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114098"},"PeriodicalIF":1.9,"publicationDate":"2025-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143877038","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":"Design and pre-test analyses of an integral thermal–hydraulic facility for a prismatic gas-cooled micro reactor","authors":"Zheng Huang ,&nbsp;Miaoxin Jiao","doi":"10.1016/j.nucengdes.2025.114097","DOIUrl":"10.1016/j.nucengdes.2025.114097","url":null,"abstract":"<div><div>To support the R&amp;D of the novel gas-cooled micro reactor (GMR), an integral thermal–hydraulic test facility (named INTALI) is designed to explore the key thermal–hydraulic phenomena specific to the GMR, thereby providing necessary data to validate computer codes for thermal–hydraulic and accident transient analysis. This paper presents a preliminary design of the INTALI facility, experimental methodology, and pre-test analyses. The INTALI facility consists of a primary loop operating at the prototypical temperature and pressure and a test section containing a scaled-down simulated reactor and a passive core cooling system (PCCS). Steady-state and transient tests will be carried out, which correspond to the normal operation and the pressurized loss of forced coolant (PLOFC) accident condition of the GMR, respectively. The experiment is mainly to investigate: (i) the coupling between the reactor and the PCCS, especially during the PLOFC, (ii) the operational characteristics of the PCCS and the energy distribution, and (iii) potential thermal stratification in the gravitational direction caused by the natural circulation in the PCCS. The pre-test analyses of the experiment were performed by CFD simulations using the COMSOL Multiphysics software. The predicted 3D distributions of the temperature and velocity fields for both the reactor and the PCCS are used to determine the instrumentation scheme. The simulation results show that no significant vertical thermal gradient is observed on the RPV wall. The radiative heat transfer from the RPV to the PCCS insulation layer plays an important role in heat removal in addition to convection. The heat removal capability of the PCCS is significantly influenced by the RPV’s temperature during the PLOFC transient. The developed CFD model is also ready for post-test quantification and validation once the experimental data is available.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"439 ","pages":"Article 114097"},"PeriodicalIF":1.9,"publicationDate":"2025-04-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143874708","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}