Nuclear Engineering and Design最新文献

{"title":"Experimental and numerical analysis on safety condenser performance based on P1 experiments at the PKL facility","authors":"I. Gómez-García-Toraño ,&nbsp;S. Schollenberger ,&nbsp;L. Dennhardt ,&nbsp;A. Wielenberg ,&nbsp;M. Vernassière ,&nbsp;S. Buchholz ,&nbsp;O.S. Al-Yahia ,&nbsp;E. Garcia ,&nbsp;M. Polidori ,&nbsp;N. Sobecki ,&nbsp;F. Lahovský ,&nbsp;F. de-Bouet-du-Portal ,&nbsp;G. Grippo ,&nbsp;M. Montout","doi":"10.1016/j.nucengdes.2025.114016","DOIUrl":"10.1016/j.nucengdes.2025.114016","url":null,"abstract":"<div><div>Passive systems are being considered for advanced reactor designs because of their enhanced reliability against an extended loss of offsite power. Particularly, the SAfety COndenser (SACO) stands out because of its capacity of passively removing core decay heat through the steam generators by condensing steam inside a immersed heat exchanger. This article presents recent experimental data and the associated numerical calculations on the vertical straight-tube SACO installed at the PKL facility. In particular, the SACO power removal capacity has been studied within the frame of test P1.1 consisting of steady state phases A, B, C and D with varying pool liquid levels and a Core Exit Temperature of 237 °C i.e. 20 K subcooling.</div><div>Experimental results show the SACO capability to transfer its nominal power of 450 kW despite the accumulation of nitrogen in the straight tubes. Improved venting procedures of phases A2 and C2 allowed a partial removal of nitrogen from the tubes and hence, an increase of the maximum core power to keep the CET constant in comparison to their counterpart phases A and C. The accumulation of nitrogen in the tubes leads to the formation of passive zones characterised by a degraded heat transfer towards the pool and significant cool-down of the liquid film.</div><div>An important numerical work has also been conducted using the CATHARE-3, ATHLET, TRACE, RELAP-5 system thermalhydraulic codes and <span><math><mtext>Neptune_cfd</mtext></math></span>, either in standalone mode or coupled with CATHARE-2. Several approaches have been adopted in order to model the primary system, SACO pool, straight tubes, boundary and initial conditions (e.g. nitrogen content, heat losses), auxiliary components (heaters, pump cooling), which add up to the physical models when analysing discrepancies with experimental results. Generally, codes are able to predict the phenomena happening in PKL, although further efforts should be invested in the use of 3D approaches to model the pool and the improvement of condensation modelling in vertical tubes for the SACO-operating region.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114016"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817333","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":"Blast response of a scaled reinforced concrete structure with Two-Leaf cavity infill wall","authors":"Ahmet Tuğrul Toy ,&nbsp;Onur Onat ,&nbsp;Barış Sevim","doi":"10.1016/j.nucengdes.2025.114055","DOIUrl":"10.1016/j.nucengdes.2025.114055","url":null,"abstract":"<div><div>Nuclear power plants, composed of boiler houses, reactors, and other facilities, operate at a high risk of explosion. Engineers design boiler hoses and other facilities to withstand dynamic loads like earthquakes, machine vibrations, wind, and blast loads. However, over time, these structures may cease to meet the requirements of current codes. Therefore, it remains unclear how different materials, their orientations, and their interactions, such as masonry and reinforced concrete, will respond in the event of a blast around a nuclear power plant. Currently, this study aims to evaluate the global and local blast response of single and two-leaf cavity infill wall enclosures with reinforced concrete structures. For this purpose, a scaled structure that is exposed to a shake table experiment has been selected. Then the structural system is numerically modelled by using ANSYS-AUTODYN and calibrated based on dynamic identification tests. The explosive amount is fixed at 78 kg to facilitate comparison of two models. For blast analysis of the structural system, two different infill wall typologies and three different scenarios are evaluated. The location of explosives determined the studied cases. We register the analytical blast responses in terms of the pressure, strain, and out-of-plane displacement of the infill wall. We limited the blast analyses to 3 ms. We compared the out-of-plane displacement of single and cavity infill walls with each other and with UFC 3–340-02. According to the findings, the thinner leaf in the Two Leaf Cavity Wall model protects the thicker leaf from damage.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114055"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143807823","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":"Molecular dynamics study of local structure and migration properties of LiCl-Li2O-Li molten salts based on machine-learned deep potential","authors":"Jitang Xu ,&nbsp;Yilin Wang ,&nbsp;Benlin Yao ,&nbsp;Yanhong Jia ,&nbsp;Yiqun Xiao ,&nbsp;Lin Zhang ,&nbsp;Bin Li ,&nbsp;Hui He ,&nbsp;Baohua Yue ,&nbsp;Liuming Yan","doi":"10.1016/j.nucengdes.2025.114052","DOIUrl":"10.1016/j.nucengdes.2025.114052","url":null,"abstract":"<div><div>The local structure and physical properties of LiCl-Li₂O-Li molten salt, the reaction medium for lithium thermal and electrolytic reduction, are very important for the study of spent fuel pyroprocessing process. In this work, the machine-learned deep potential (MLDP) was trained using dataset based on first-principle molecular dynamics (FPMD) and was used to predict the changes in the physical properties of molten LiCl with the addition of different concentrations of Li₂O and Li between 923 K and 1323 K. Deep potential molecular dynamics (DPMD) calculations were performed for properties including shear viscosity, electrical conductivity, thermal conductivity, and specific heat capacity. It was revealed that the addition of Li significantly reduces the diffusion activation energies (<span><math><mrow><msub><mi>E</mi><mi>a</mi></msub></mrow></math></span>) of Li⁺ and Cl^- in the molten salt. By comparison with the experimental data of pure LiCl, it can be concluded that the MLDP can describe the inter-atomic interactions of molten salt correctly, overcome the problem of missing potential parameters in the classical inter-atomic empirical potentials. Finally, DPMD allows to simulate large systems with comparable accuracy of FPMD, thus provide theoretical guidance for the optimization of the pyroprocessing technology.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114052"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817332","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 of novel refractory equiatomic multi-principal elemental alloys based on Mo-Nb-Ti system for Gen IV reactor applications","authors":"Anilas Karimpilakkal ,&nbsp;Joseph W. Newkirk ,&nbsp;Jason L. Schulthess ,&nbsp;Frank Liou ,&nbsp;Visharad Jalan ,&nbsp;Haiming Wen","doi":"10.1016/j.nucengdes.2025.114050","DOIUrl":"10.1016/j.nucengdes.2025.114050","url":null,"abstract":"<div><div>Excellent irradiation damage resistance demonstrated by multi-principal elemental alloys (MPEAs) has sparked significant interest among researchers, prompting exploration into their vast compositional space, to validate their suitability for nuclear applications. A combined approach of thermodynamic and empirical parameters calculations alongside CALPHAD (CALculation of PHAse Diagrams) for phase formation predictions enable high-throughput material selection for sophisticated applications like nuclear, overcoming laborious and time-consuming experiments. Key thermodynamic and empirical parameters for eight novel equiatomic MPEAs, based on seven low thermal neutron cross section refractory elements, for predicting phase formation were calculated, and equilibrium and non-equilibrium simulations in CALPHAD were employed to comprehensively evaluate the systems. Pseudo binary phase diagram simulations showed that Zr, V or equiatomic CrV additions to the base MoNbTi alloy (MoNbTi-Zr, MoNbTi-V and MoNbTi-CrV alloys) favor the formation of isomorphous body-centered cubic (BCC) phase at high temperatures, while Cr, Al, equiatomic ZrV, or equiatomic CrAl additions (MoNbTi-Cr, MoNbTi-Al, MoNbTi-ZrV or MoNbTi-CrAl alloys) limit the solubility of them. Equilibrium CALPHAD simulations at 750 °C were consistent with XRD results on MoNbTi, MoNbTiZr and MoNbTiCr alloys, and partially for others. Notably, elemental segregation observed in the backscattered electron (BSE) scanning electron microscopy (SEM) images of the alloys was accurately simulated through non-equilibrium Scheil solidification calculations in CALPHAD, further verified by experiments. The precipitation of TiCr₂ Laves phase in Cr containing MoNbTiCr and MoNbTiCrAl was accurately predicted while discrepancies were noted in MoNbTiCrV. The equilibrium simulations also provided insights into phase compositions at specific temperatures offering a pathway for tailoring the desired microstructure and properties of these systems. Empirical parameters calculations successfully predicted random solid solution in the base MoNbTi alloy, and with an exception in MoNbTiV and MoNbTiAl, predicted intermetallic precipitation in the rest, especially, Laves phase precipitation in Cr containing alloys.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114050"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817337","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":"Bamboo-SFuel: A nuclide composition evaluation code for PWR spent fuel in NECP-Bamboo","authors":"Yunzhao Li ,&nbsp;Songzhe Wang ,&nbsp;Yuancheng Zhou ,&nbsp;Senhan Yang ,&nbsp;Ruizhi Shao ,&nbsp;Liangzhi Cao ,&nbsp;Tian Chen ,&nbsp;Miao Yu ,&nbsp;Zeng Shao ,&nbsp;Guoming Liu","doi":"10.1016/j.nucengdes.2025.114047","DOIUrl":"10.1016/j.nucengdes.2025.114047","url":null,"abstract":"<div><div>To accurately obtain the nuclide composition in spent fuel, a nuclide composition calculation program named Bamboo-SFuel has been developed as a new component in the PWR-core analysis software NECP-Bamboo. It can switch between different depletion data libraries, including one with 233 nuclides in the lattice code Bamboo-Lattice, the one with 1547 nuclides, or the one with 3838 nuclides. The shortcomings of existing domestic procedures, such as incomplete nuclide types in the data library and incomplete simulation of the irradiation process, limit the reliability and economy of the safety analysis of spent fuel reprocessing. The nuclide composition was quantitatively verified and analyzed using the measured data of Post-Irradiation Experiments (PIE). The source term calculation accuracy of the program was verified by comparing with the calculation results of the SCALE package, and the impact of different depletion data libraries on the nuclide composition and source term calculation results was also explored. Encouraging conclusions have been drawn from the numerical results. (1) Bamboo-SFuel can accurately analyze the radionuclide composition and radioactive source term of spent fuel assemblies of commercial PWR under different irradiation conditions. (2) The depletion data library with different numbers of nuclides has little influence on the calculation results of important nuclide composition, but has a great influence on the total radioactive source term calculation results. (3) Based on the built-in depletion data library containing 1547 nuclides, this program can simultaneously provide reliable radioactive source analysis of spent fuel and important nuclide composition concerned by radiation safety analysis.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114047"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817331","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":"Performance analysis of Dowtherm A heat pipe with internal vapor monitoring","authors":"Mitchell Stephenson ,&nbsp;Trevor Melsheimer ,&nbsp;Joseph Seo ,&nbsp;Abdulbasit Aloufi ,&nbsp;Hansol Kim ,&nbsp;Yassin A. Hassan","doi":"10.1016/j.nucengdes.2025.114049","DOIUrl":"10.1016/j.nucengdes.2025.114049","url":null,"abstract":"<div><div>Medium-temperature heat pipes, operating in the 200–600 °C range, find widespread application in sectors such as nuclear microreactors, solar energy collectors, thermal energy storage, and space. Efficient, passive heat transfer devices, like heat pipes, are essential for power systems operating in this temperature range. Despite such a broad range, traditional working fluids for heat pipes in the medium-temperature regime frequently underperform, prompting the need for more research into these working fluids. Dowtherm A is attractive for its chemical compatibility with heat pipe materials, low toxicity, low flammability, and adequate thermal–hydraulic properties, things that cannot be said for most medium-temperature heat pipe working fluids. This experimental study investigates the performance of Dowtherm A as a medium-temperature heat pipe working fluid, using internal and external measurements to quantify the heat transport in the heat pipe. A 25.4 mm outer diameter, 316 stainless steel tube was used for the heat pipe testing. Ten wraps of 100 × 100 (100 openings per inch) 316 stainless steel screen mesh were used as the wick, with a sliding fit and no annular gap. A fill ratio of 103 % of the total wick void volume was used. An air jacket was attached to the condenser of the heat pipe for cooling. Internal and external temperature measurement was performed, utilizing optical fiber distributed temperature sensing and conventional thermocouples, respectively. All tests conducted were in the horizontal orientation. The test matrix consisted of three different cooling conditions, controlled by changing the flow rate of air in the jacket over the condenser, with multiple power levels for each cooling condition. It was found that the thermal resistance of the heat pipe is not influenced directly by the cooling flow rate but is instead linked to the operating temperature. A minimum thermal resistance of 0.58 °C/W was achieved at the highest operating temperature tested of 274 °C. This corresponds to a maximum effective thermal conductivity of 2300  W/m·K. This finding agrees with values from previous studies. Internal vapor temperature measurements determined the active condenser length, where vapor condenses—a useful tool in heat pipe design. The capillary limit, which governs power transport in heat pipes, was exceeded in all tests without dryout, suggesting Dowtherm A outperformed expectations. This finding questions the soundness of the commonly used theoretical capillary limit, as applied for organic fluids such as Dowtherm A. Collectively, these findings highlight Dowtherm A’s viability for use in medium-temperature heat pipes, offering improved efficiency and operational safety in diverse energy systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114049"},"PeriodicalIF":1.9,"publicationDate":"2025-04-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143817334","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":"Assessment of a novel fuel block and core arrangement for use in a nitrogen-cooled direct cycle high temperature gas-cooled reactor","authors":"Jeremy Henry Owston","doi":"10.1016/j.nucengdes.2025.114040","DOIUrl":"10.1016/j.nucengdes.2025.114040","url":null,"abstract":"<div><div>This paper investigates the thermal performance of a new fuel block design utilising annular compacts supported in counterbored holes through a coupled thermohydraulic and neutron transport study of a prospective High Temperature Gas cooled Reactor (HTGR) core. The paper highlights the design freedom afforded by a fuel block design which permits irregular spacings of fuel channels without impacting the heat transport to the coolant channel. The approach of optimising the moderating ratio through variable fuel spacings can flatten the radial thermal flux profile, achieving thermal peaking flux factors of less than 1.15 for cores studied in this paper. Flattening the radial thermal flux is shown to minimise variations in coolant outlet temperatures and therefore significantly reduce peak fuel temperatures in the core.</div><div>Burn-up studies of the core demonstrate the benefits of a radially optimised thermal flux profile by demonstrating insensitivity to the fuel burnup of the power profile within the core. This insensitivity results in consistent peak coolant outlet temperatures and small variations in peak fuel temperature over the course of the core life.</div><div>The paper also demonstrates the design flexibility offered by using variable diameter coolant channel displacer rods within the centre of each fuel channel to enhance heat transfer, whilst also balancing the flow distribution within the core. Specifically, the approach of utilising variable displacer rod diameters to match local coolant mass flow with fuel column power is shown to reduce peak fuel temperatures where significant power peaking factors exist.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114040"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799001","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":"Thermo-economic-environmental analysis and performance-based Pareto optimization of a floating nuclear power plant","authors":"Masoud Nasouri ,&nbsp;Navid Delgarm","doi":"10.1016/j.nucengdes.2025.114013","DOIUrl":"10.1016/j.nucengdes.2025.114013","url":null,"abstract":"&lt;div&gt;&lt;div&gt;This paper presents a comprehensive thermo-economic-environmental analysis and performance-based Pareto optimization of a Floating Small Modular Reactor Power Plant (F-SMRP) along the Bushehr coast (Iran), designed to meet regional electricity demands. Innovative ideas are employed to predict the thermodynamic properties of the F-SMRP using an artificial neural network. Upon model verification, a detailed 4E (energy, exergy, exergoeconomic, and environmental economics) analysis is conducted. Further, performance optimization is carried out targeting key metrics such as exergy efficiency (&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;II&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;), the total capital cost rate (&lt;span&gt;&lt;math&gt;&lt;msubsup&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mrow&gt;&lt;mi&gt;tot&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;/mrow&gt;&lt;/msubsup&gt;&lt;/math&gt;&lt;/span&gt;), and the total product exergy cost rate (&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt;) using the artificial bee colony algorithm. The final optimal configuration, referred to as the Optimized F-SMRP (OF-SMRP), is determined through the analytic hierarchy process decision-making. The results demonstrate that F-SMRP achieves &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; and &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;II&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; of 29.9 % and 64.12 %, respectively, with corresponding &lt;span&gt;&lt;math&gt;&lt;msubsup&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mrow&gt;&lt;mi&gt;tot&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;/mrow&gt;&lt;/msubsup&gt;&lt;/math&gt;&lt;/span&gt; and &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mrow&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; of $331.8/hour and $6191.8/hour. In contrast, OF-SMRP exhibits notable improvements across all metrics compared to the F-SMRP. The &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; reaches 33.2 %, showing a significant increase of 11 %. Also, &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mover&gt;&lt;mi&gt;W&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mi&gt;e&lt;/mi&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; is 34.81 MW, showcasing an increase of 2.11 MW. Similarly, the &lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;II&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; improves to 68.1 %, representing a 3.97 % gain. Despite a moderate rise in &lt;span&gt;&lt;math&gt;&lt;msubsup&gt;&lt;mover&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mo&gt;̇&lt;/mo&gt;&lt;/mover&gt;&lt;mrow&gt;&lt;mi&gt;tot&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mspace&gt;&lt;/mspace&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mo&gt;-&lt;/mo&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;mrow&gt;&lt;mi&gt;I&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;O&lt;/mi&gt;&lt;mo&gt;&amp;&lt;/mo&gt;&lt;mi&gt;M&lt;/mi&gt;&lt;/mrow&gt;&lt;/msubsup","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114013"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791708","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":"Criticality and sensitivity analysis of VVER-1000 mock-up with SCALE and MCNP5 code using ENDF/B-VII.1 nuclear data library","authors":"Mohammad Omar Faruk ,&nbsp;Mohammad Abdul Motalab ,&nbsp;Mohammad Sayem Mahmood ,&nbsp;Gil Soo Lee","doi":"10.1016/j.nucengdes.2025.114015","DOIUrl":"10.1016/j.nucengdes.2025.114015","url":null,"abstract":"<div><div>Accurate analysis of reactor criticality is essential for reactor design and safety assessments. This paper conducts a criticality study of a VVER-1000 mock-up benchmark experiment, which was performed at the LR-0 research reactor operated by the Research Center Rez in the Czech Republic. Benchmark calculations are performed using two Monte Carlo codes – SCALE (KENO-VI) and MCNP5 – utilizing the ENDF/B-VII.1 continuous-energy nuclear data library for criticality calculations. The mock-up was examined under six different critical configurations by varying coolant levels and boric acid concentrations. This paper provides a comparative analysis of the results from SCALE (KENO-VI) and MCNP5 to assess the suitability of SCALE (KENO-VI) as a verification tool in the regulatory process, with MCNP5 as the reference code. Additionally, the research work also investigates the sensitivity of various reactor system parameters’ uncertainty, highlighting their significant impact on criticality result, which could potentially lead to overly conservative safety margin. The study focuses uncertainty on five key technological parameters: fuel assembly pitch, fuel cladding thickness, fuel density, fuel enrichment and boric acid concentration. A comprehensive analysis of these uncertainties, along with an assessment of their sensitivity to the criticality results, is provided in this study.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":""},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791133","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":"Investigating the nexus between radiolysis using spent nuclear fuel and hydrogen production, with environmental safety considerations – A literature review","authors":"Ephraim Bonah Agyekum ,&nbsp;Flavio Odoi-Yorke ,&nbsp;Mustafa Abdullah ,&nbsp;Prangon Chowdhury","doi":"10.1016/j.nucengdes.2025.114048","DOIUrl":"10.1016/j.nucengdes.2025.114048","url":null,"abstract":"<div><div>Given the detrimental consequences of climate change on the environment, hydrogen appears to be one of the solutions to possible decarbonization. Although it is a potential option, the subject of hydrogen production utilizing nuclear spent fuel (SNF) through water radiolysis has not received much attention to explore its potential use for large-scale hydrogen production. This review, therefore, presents an overview of research work on the use of SNF’s ionizing radiation for the production of hydrogen and its potential impact on environmental safety. This paper investigates the advantages and difficulties of hydrogen generation in SNF storage system by utilizing a bibliometric and systematic review technique to analyse previous research works. The research themes were classified into motor, niche, emerging/declining, and basic themes. Some important themes that were found to be central to the topic included radiation shielding, hydrogen production, and environmental sustainability; life cycle assessment; renewable energy integration; nuclear waste management; hydrogen storage; and modular reactors. The study identified some potential research gaps and provided some recommendations for future research. This includes the improvement in the hydrogen detection systems, hydrogen collection and ventilation, containment of radioactive isotopes, understanding radiolysis impacts, and the development of purification and storage methods for hydrogen from spent nuclear fuel. The outcome of this study is expected to shape future research on the subject matter and serve as the foundation for deeper understanding of radiolysis for commercial-scale hydrogen production.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114048"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143791716","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}