Nuclear Engineering and Design最新文献

{"title":"Cementitious mortar for Intermediate and Low-Level Radioactive waste confinement: Matrix optimization and leaching","authors":"Domenico Rosa ,&nbsp;Francesco Rizzo ,&nbsp;Teresa Mangialardi ,&nbsp;Franco Medici ,&nbsp;Gianfranco Caruso ,&nbsp;Luca Di Palma ,&nbsp;Fabio Giannetti","doi":"10.1016/j.nucengdes.2025.114165","DOIUrl":"10.1016/j.nucengdes.2025.114165","url":null,"abstract":"<div><div>A cement-based mortar prepared utilizing a pozzolanic cement (CEM IV/A 42.5 R type conforming to European Standard EN 197–1) was optimized for the containment of Intermediate and Low-Level Radioactive Waste without the use of expensive additives. Surrogate radionuclides, including lithium (Li), cesium (Cs), cobalt (Co), and lead (Pb) salts, were incorporated to simulate real waste streams.</div><div>The experiments focused on optimizing the water-to-cement ratio, the sand-to-cement ratio of the mortar, and assessing the effects of simulated radionuclide addition on its properties. At the investigated level of adding radionuclides (2.5 mmol/kg), the mortar maintained a compressive strength of approximately 54 N/mm², while water absorption and workability remained unchanged with respect to the reference. However, thermogravimetric analysis and Fourier-transform infrared spectroscopy indicated that radionuclides interfered with the hydration reaction.</div><div>Leaching tests on hardened monolith specimens demonstrated varying radionuclide mobilities, with the following order: Li &gt; Cs &gt; Co &gt; Pb. In all cases, the leachability index was greater than 8, confirming effective containment. Diffusion was identified as the controlling mechanism for all simulated radionuclides except Pb, for which wash-off dominated the release process. These results confirm that CEM IV/A based mortar is an effective material for immobilizing radionuclides, maintaining mechanical integrity while limiting radionuclide migration.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114165"},"PeriodicalIF":1.9,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138252","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":"Research progress on tritium protective materials","authors":"Caixia Miao ,&nbsp;Xiaohui Du ,&nbsp;Ke Ding ,&nbsp;Dehong Wang ,&nbsp;Hailei Lv ,&nbsp;Hongchen Han ,&nbsp;Yongshuai Xie","doi":"10.1016/j.nucengdes.2025.114169","DOIUrl":"10.1016/j.nucengdes.2025.114169","url":null,"abstract":"<div><div>With the development of advanced nuclear energy, tritium radiation protection and safe storage of tritium-containing waste have received extensive attention. Reliable tritium protection materials are the premise and basic guarantee for the secure development of tritium. This work summarizes the research progress of corresponding protective materials around key links such as personnel tritium protection, protection of tritium-related equipment and facilities, safe disposal of tritium-containing wastewater and waste gas, and storage of tritium-containing waste. It analyzes the future development direction of various materials, aiming to provide a reference for the safety protection of tritium-related personnel and equipment and the development of related materials.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114169"},"PeriodicalIF":1.9,"publicationDate":"2025-05-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144124788","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":"Dynamic exergy optimization of feedwater heater start-up process in nuclear power plant using optimal control method","authors":"Zhijiang Zhang,&nbsp;Zhaofei Tian","doi":"10.1016/j.nucengdes.2025.114163","DOIUrl":"10.1016/j.nucengdes.2025.114163","url":null,"abstract":"<div><div>The operational optimization of energy systems is of great significance for the sustainable development of energy, as well as for achieving carbon peaking and carbon neutrality. This research took the feedwater heater in a nuclear power unit as the research object, developed a thermo-hydraulic model applicable to the optimization. The optimal control algorithm, which is widely used in chemical and energy systems, has been innovatively applied to nuclear power units and combined with the dynamic exergy analysis method to optimize the start-up process. The results show that although the time optimal control shortens the startup time by 19.69%, the cumulative exergy loss increases by 1 time. The optimal control with the highest exergy efficiency as the goal reduced the start-up time by 19.23%, and the accumulated exergy destruction decreased by 25.14%. This study greatly reduced the startup time of feedwater heater and improved the process exergy efficiency. This study provided important guidance for the dynamic exergy optimization of energy system start-up processes and addressed the long-standing limitation of exergy analysis being confined to steady-state conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114163"},"PeriodicalIF":1.9,"publicationDate":"2025-05-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144116997","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":"Comprehensive evaluation of nuclear data library variations and TRISO distribution methods on neutronic properties of HTR-10 pebble-bed fuel using OpenMC","authors":"Ahmad Muzaki Mabruri ,&nbsp;Nuri Trianti ,&nbsp;Zaki Su’ud ,&nbsp;Efrizon Umar ,&nbsp;Ratna Dewi Syarifah","doi":"10.1016/j.nucengdes.2025.114168","DOIUrl":"10.1016/j.nucengdes.2025.114168","url":null,"abstract":"<div><div>The HTR-10 reactor requires highly complex modeling, particularly in pebble distribution within the core and TRISO particle distribution within the pebbles. This complexity significantly impacts the suitability of nuclear data library selection and the accuracy of model simplifications to enhance computational efficiency. While most related studies focused on <em>k-</em>eigenvalue calculations, they often overlooked depletion calculations, which were essential for determining reactor operational lifetime. This study comprehensively evaluated the impact of four nuclear data libraries (ENDF/B-VIII.0, ENDF/B-VII.1, JEFF-3.2, and JEFF-3.3) and three TRISO distribution methods (random lattice, hexagonal lattice, and rectangular lattice) on the neutronic characteristics of HTR-10 pebble-bed fuel, including <em>k</em>-eigenvalue, material depletion, and the dominant physical properties influencing these parameters. The results indicated that all four data libraries provided consistent criticality calculations, but significant differences were observed in material depletion, particularly between ENDF/B-VIII.0 and ENDF/B-VII.1, with relative difference remaining below 1.00 %. The geometric simplification aspect that was implemented resulted in a sixfold reduction in computation time for <em>k-</em>eigenvalue calculations and a 43-fold reduction for depletion calculations, without compromising accuracy or precision, maintaining differences below 1.00 % for k-eigen, material depletion, and other physical parameters. Notable discrepancies between simplified and random models were primarily observed in reaction rate distributions, yet these did not substantially affect total reaction rate calculations, with relative errors remaining under 1.5 × 10⁻³. These findings highlighted that simplifying TRISO distribution using hexagonal lattice and rectangular lattice significantly enhanced the efficiency of HTR-10 pebble-bed fuel analysis without causing significant deviations in accuracy or precision.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114168"},"PeriodicalIF":1.9,"publicationDate":"2025-05-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144124287","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":"Review on corrosion-related unidentified deposit of pressurized water reactors","authors":"Hui He ,&nbsp;Yan Liu ,&nbsp;Shiwei Wang ,&nbsp;Tengfei Zhang ,&nbsp;Xiang Chai ,&nbsp;Xiaojing Liu","doi":"10.1016/j.nucengdes.2025.114095","DOIUrl":"10.1016/j.nucengdes.2025.114095","url":null,"abstract":"<div><div>Corrosion-related unidentified deposit (CRUD) on the fuel cladding is detrimental for nuclear reactor safety and economy. This paper provides a comprehensive review of published articles and disclosed technical reports concerning CRUD of pressurized water reactors in terms of its formation, properties, and consequences, which is intended to better understand and predict this complicated issue. The formation of CRUD is a physicochemical process involving the thermodynamics of soluble species and kinetics of insoluble particles. Changes to the local conditions, such as primary water chemistry, thermal hydraulic parameters, morphology feature of metal surface, affect the properties of CRUD including ingredients and multi-scale structures. Abnormal axial power distribution occurs due to the neutron absorption by boron-containing compounds in porous CRUD, which is related to the capillary flow, heat transfer, solute transport as well as chemical reactions. This paper also discusses the potential strategies to mitigate the CRUD deposits, such as water chemistry control, chemical purification, ultrasonic cleaning and coating. The most attractive candidate for CRUD prevention at current stage is zinc injection approach under elaborate control of water chemistry, coating approach will be another attractive candidate for CRUD prevention as development of accident tolerant fuels (ATF) from both scientific and manufactural perspectives.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114095"},"PeriodicalIF":1.9,"publicationDate":"2025-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144116955","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":"Uncertainty quantification application on the Phebus FPT1 using the coupling of MELCOR and DAKOTA in a Python environment/architecture","authors":"Giuseppe Agnello ,&nbsp;Mattia Massone ,&nbsp;Fulvio Mascari","doi":"10.1016/j.nucengdes.2025.114135","DOIUrl":"10.1016/j.nucengdes.2025.114135","url":null,"abstract":"<div><div>During the last few years, the international nuclear scientific community involved in the development of deterministic safety analyses in the Severe Accident (SA) domain has focused its interest on the analysis of methodologies for the uncertainty quantification using the state-of-art integral SA codes (e.g. ASTEC, MAAP, MELCOR, etc.). Within this framework, the H2020 “Management and Uncertainty of Severe Accidents” (MUSA) project, coordinated by CIEMAT (Spain), aimed to establish a harmonized approach for the application of uncertainty quantification methodologies to SAs. Along with the MUSA Working Package named “Application of UQ Methods against Integral Experiments”, coordinated by ENEA (Italy), the uncertainty analysis methodology has been applied to the Phebus FPT1 experiment. To develop the reference base case, the integral code MELCOR 2.2 has been selected, and it has been coupled with the uncertainty tool DAKOTA to develop the uncertainty analysis. The study has been focused on the degradation and beginning of the aerosol phase of the test, considering the MELCOR aerosol miscellaneous constants as uncertain input parameters and the maximum value of the total mass of aerosol in suspension in the containment atmosphere as Figure of Merit (FOM). The uncertainty analysis permitted a first estimation of the uncertainty bandwidth of the FOM, comparing it with the experimental and reference case values. Among the uncertain input parameters, the dynamic shape factor and the agglomeration shape factor present, respectively, a moderate and significant statistical correlation with the FOM.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114135"},"PeriodicalIF":1.9,"publicationDate":"2025-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144106385","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":"Analysis of heat pipe cooled reactor startup characteristics by multi-physics coupling simulation based on MOOSE framework","authors":"Weixiang Wang ,&nbsp;Kefan Zhang ,&nbsp;Sifan Dong ,&nbsp;Shuai Wang ,&nbsp;Hongli Chen","doi":"10.1016/j.nucengdes.2025.114164","DOIUrl":"10.1016/j.nucengdes.2025.114164","url":null,"abstract":"<div><div>A Heat pipe reactor utilizes heat pipes to transfer fission heat from the reactor core, providing a high level of passive safety and thus representing a promising application for micro-reactors. The startup process of a heat pipe reactor is characterized by a wide range of temperature variations and a dominant expansion reactivity feedback, while also necessitating consideration of the startup issues of the frozen alkali metal heat pipes, which adds significant complexity. To analyze the characteristics of the startup process of the heat pipe reactor, this study couples the MOOSE framework with the two-dimensional high-temperature heat pipe numerical simulation code KMC-HPs, to conduct a high-precision multi-physics coupling simulation study of the nuclear-thermal–mechanical dynamics, and the accuracy and feasibility of this coupling method were validated through simulations of the ground prototype reactor KRUSTY. The analysis indicates that the heat transfer characteristics during the startup process of the heat pipes determine the temperature distribution within the reactor core. Initially, the heat transfer capability of the heat pipes is limited, resulting in a significant temperature difference. However, as the temperature rises and continuous vapor flow occurs in the evaporation section of the heat pipes, the heat transfer capability improves, leading to a more uniform temperature increase across the core with a smaller temperature difference. Mechanical analysis reveals that large thermal stresses occur at the edges of the contact surfaces between the heat pipes and the core, although these stresses remain well below the material’s yield limits. This paper elucidates the multi-physical characteristics of the heat pipe reactor startup process from a numerical simulation perspective, and the developed coupling framework provides high-precision validation for the conceptual design of heat pipe reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114164"},"PeriodicalIF":1.9,"publicationDate":"2025-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144116998","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 heat transfer characteristics of debris bed remelting process","authors":"Lijun Jian ,&nbsp;Xiao Zeng ,&nbsp;Peng Yu ,&nbsp;Yidan Yuan","doi":"10.1016/j.nucengdes.2025.114173","DOIUrl":"10.1016/j.nucengdes.2025.114173","url":null,"abstract":"<div><div>The researches of IVR (In-Vessel Retention) primarily focuse on natural convection in steady-state molten pools. However, the dynamic process from particulate debris bed melting to molten pool is also critical. The debris beds remelting process may exhibit intermediate states that potentially exceed steady-state thermal loads, while simultaneously serving as the initial condition for steady-state molten pools. This study conducted experimental investigations on single-component and two-component debris bed remelting processes using the COREM-HT experimental facility. The results included phenomenological characteristics, temperature evolution and heat flux distribution. In both experiments, no transient heat flux exceeding the final state was observed. Notably, two crust layer observed on the wall differed from conventional steady-state molten pool configurations with single oxide crust. These findings offer significant references for IVR research.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114173"},"PeriodicalIF":1.9,"publicationDate":"2025-05-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144106384","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":"A comparative analysis of the flow in a rod bundle with two guide-tubes and mixing grids, URANS simulation and Wall-Modeled LES of the ALAIN experiment","authors":"Antoine Michel,&nbsp;André Bergeron,&nbsp;Maria Adela Puscas","doi":"10.1016/j.nucengdes.2025.114115","DOIUrl":"10.1016/j.nucengdes.2025.114115","url":null,"abstract":"<div><div>The flow characteristics in a 5 × 5 rod bundle representative of ALAIN experiment is investigated using Unsteady Reynolds-Averaged Navier–Stokes and Wall-Modeled Large Eddy Simulation methodologies. A review of the numerical methodology is first presented to highlight the impact of mesh resolution and computational domain size on simulation outcomes. Experimental measurements of the velocity along two traverses of the rod-bundle lattice are then employed as reference data to evaluate the accuracy of computational fluid dynamics simulations. The average error on the velocity statistics reveals substantial discrepancies between the numerical results and experimental measurements. The disturbance induced by the mixing grids and the presence of guide-tubes represent a challenging feature to reproduce numerically, both at close and long distance from the grids. Computation on the error made on the velocity statistics reveals that Wall-Modeled Large Eddy Simulation perform better than the <span><math><mrow><mi>k</mi><mo>−</mo><mi>ϵ</mi></mrow></math></span> URANS model to predict the axial velocity at close distance from the grid and the transverse flow at large distance from the grid, but comes at a much higher computational cost. The present analysis enhances the understanding of flow dynamics in rod bundles and informs the selection of appropriate turbulence models for future simulations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114115"},"PeriodicalIF":1.9,"publicationDate":"2025-05-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144098331","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 GAMA-SHP microreactor for outer space application: Neutronic study","authors":"Andang Widi Harto,&nbsp;Kusnanto,&nbsp;Alexander Agung,&nbsp;M.Yayan Adi Putra,&nbsp;Dani Abdul Aziz N.","doi":"10.1016/j.nucengdes.2025.114167","DOIUrl":"10.1016/j.nucengdes.2025.114167","url":null,"abstract":"<div><div>Outer space researchs and explorations require reliable electrical supplies. For outer space explorations far from the sun, nuclear reactors become reliable for electricity generation. The space reactor must be simple, safe, and long live. GAMA Space Heat Pipe (GAMA-SHP) is a nuclear reactor designed to generate electricity for space applications with 1.2 MWe output power, fueled with ThF₄ and UF₄ mixture. The reactor is cylindrical shape with 100 cm diameter, 100 cm height and has 20 cm radial and axial graphite reflectors. Calculations using Open MC software have been performed to calculate reactor criticality, temperature reactivity coefficient, void reactivity coefficient, optimation of control rod position, control rod worth, power distribution and burn up. The GAMA-SHP cannot achieve critical condition without reflector. With reflector, GAMA-SHP can achieve critical condition at more than 90 % mole fraction of UF₄. The reactivity difference with and without the reflector is 6648 pcm. The effective fuel temperature reactivity coefficient is <span><math><mrow><mo>-</mo><mn>4.790</mn></mrow></math></span> pcm/K. The fuel void reactivity coefficient is <span><math><mrow><mo>-</mo><mn>297.52</mn></mrow></math></span> pcm/%. With negative fuel temperature and fuel void reactivity coefficients, GAMA-SHP is inherently safe. The control rod optimal configuration is one central rod at the axial axis of the reactor (Ring 0) and six peripheral rods at Ring 6. At the beginning of the lifespan of the reactor, the criticality value when all rods are fully inserted at the optimal configuration is 0.937150 (reactor reactivity of <span><math><mrow><mo>-</mo><mn>6707</mn></mrow></math></span> pcm). With a thermal power of 5 MWth, the GAMA-SHP can maintain the critical condition for 10 years.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114167"},"PeriodicalIF":1.9,"publicationDate":"2025-05-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144098380","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}