Annals of Nuclear Energy最新文献

{"title":"Melting of Fe and Gd oxide loaded geopolymers with nuclear fuel for ex-vessel core catcher systems","authors":"Bence Mészáros ,&nbsp;Jan Hrbek ,&nbsp;Mykhaylo Paukov ,&nbsp;Tomáš Černoušek ,&nbsp;Jan Sklenka ,&nbsp;Zbyněk Černý ,&nbsp;Pavlína Rosypal ,&nbsp;Václav Tyrpekl","doi":"10.1016/j.anucene.2025.111602","DOIUrl":"10.1016/j.anucene.2025.111602","url":null,"abstract":"<div><div>Geopolymers have demonstrated a significant potential in various fields of the nuclear industry. They can serve as enhanced sacrificial materials offering increased safety and mitigating consequences in a hypothetical severe nuclear accident. Additionally, geopolymers show potential use as immobilization matrixes for radioactive waste disposal due to their advantageous properties and polymeric structures. Their properties, including high gamma radiation resistance, low water content, and the ability to incorporate various elements and species to customize their physico-chemical properties, suggest that geopolymers may be applied for various systems, e.g. sacrificial materials for core catchers of current or future plant concepts. However, further research is required to fully understand the interaction between geopolymers and corium melts, to describe their physico-chemical properties at extreme temperatures, high-temperature phase behaviour, etc. In present manuscript, we focused on geopolymers with embedded Gd₂O₃ and Fe₂O₃ as neutron absorber and functional melt modifier, respectively. Such composite could be a convenient sacrificial material for ex-vessel core catcher systems. We conducted two midscale melting experiments of geopolymer mixtures with prototypic corium to estimate the solidus and liquidus of the mixture and asses the phase behaviour. The mixture of geopolymer and prototypic corium showed similarities to previously described corium-concrete systems with solidus temperature around 1500 °C and liquidus around 1900 °C. A miscibility gap was present in the system between the oxide and iron melt. Iron melt solidified into a large block, but to microdroplets as well. Three distinct phases (silicate-rich, uranium-rich and metallic) were observed in the ingots. The geopolymer materials appeared to have prospective properties from melting and phase point of view, thus deserve further attention in the severe accident R&amp;D.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111602"},"PeriodicalIF":1.9,"publicationDate":"2025-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144523362","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":"On-The-Fly multigroup cross section generation method with continuous energy accuracy","authors":"Farzad Rahnema,&nbsp;Dingkang Zhang","doi":"10.1016/j.anucene.2025.111640","DOIUrl":"10.1016/j.anucene.2025.111640","url":null,"abstract":"<div><div>In this paper, an on-the-fly multigroup cross section generation method is developed and implemented into the coarse mesh transport code COMET to account for the effects of spectral change and surrounding environment in reactor core calculations. The method consists of (1) a continuous energy stochastic response reaction rate generation module and (2) a deterministic multigroup cross section updating module. The first module is developed by extending the continuous energy response function generator for COMET to solve a set of local fixed-source transport problems with incident flux boundary conditions (or surface sources) and to generate a library of various response reaction rates. The second module constructs the phase space (continuous in space, angle, and energy) flux distribution in the regions of interest as a superposition of elements from the pre-computed library and then generates the multigroup cross sections on-the-fly. The method is benchmarked against the multigroup cross sections directly by continuous energy Monte Carlo in a set of stylized Very High Temperature Reactor (VHTR) single fuel block and whole core benchmark problems. Benchmark results demonstrate that the new method can achieve accuracy close (i.e., within two standard deviation of stochastic uncertainty) to that of continuous energy Monte Carlo while having on-the-fly computational speed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111640"},"PeriodicalIF":1.9,"publicationDate":"2025-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144329905","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 molten salt tritium transport experiment: A pumped fluoride salt loop for hydrogen isotope experimentation","authors":"Thomas F. Fuerst,&nbsp;Anthony G. Bowers,&nbsp;Hanns A. Gietl,&nbsp;Nicole L. France,&nbsp;L.Shayne Loftus,&nbsp;Adriaan A. Riet,&nbsp;Matthew D. Eklund,&nbsp;Chase N. Taylor,&nbsp;Masashi Shimada","doi":"10.1016/j.anucene.2025.111659","DOIUrl":"10.1016/j.anucene.2025.111659","url":null,"abstract":"<div><div>Molten salt reactors and fusion reactors propose to use molten salt as coolants and breeder blanket materials. Tritium, however, poses safety concerns in both reactor types due to its ability to permeate through reactor materials creating the potential for environmental release. This article addresses the tritium transport phenomena in molten salts and presents the design and analysis of the Molten Salt Tritium Transport Experiment (MSTTE). MSTTE is a forced-convection fluoride salt loop intended to measure hydrogen isotope permeation through structural materials in a flowing salt system. Computational fluid dynamics analysis ensures fully developed salt flow in the permeation test section. MSTTE is modeled with MELCOR-TMAP to predict the permeation rate as a function of experimental variables such as source rate, salt flow rate, and salt temperature. Additionally, pressure drop analysis is conducted and finite-element analysis assesses thermal stress during loop operation to ensure the experiment’s safe design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111659"},"PeriodicalIF":1.9,"publicationDate":"2025-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144320968","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":"Semi-empirical model for resuspension of multilayer sedimentary aerosols in pipes","authors":"Peizheng Hu ,&nbsp;Zhichao Gao ,&nbsp;Yixin Huang ,&nbsp;Lili Tong ,&nbsp;Xuewu Cao","doi":"10.1016/j.anucene.2025.111652","DOIUrl":"10.1016/j.anucene.2025.111652","url":null,"abstract":"<div><div>As a critical phenomenon influencing radioactive release assessment during nuclear severe accidents, aerosol resuspension requires in-depth investigation, as it can affect the accuracy of radioactive release source term assessment. An experimental apparatus has been established and the resuspension of multi-layered sedimentary aerosols have been conducted under turbulent airflow conditions with Reynolds number ranging from 50,000 to 130,000. The experiments indicate that higher friction velocity of turbulent pipe and larger deposited particle size both increase the resuspension rate. Through mechanical fulcrum model analysis, the resuspension characteristics, which encompass the coupling effects of airflow characteristics, particle characteristics and wall characteristics, are revealed by the dimensionless particle diameter <span><math><msubsup><mi>d</mi><mi>p</mi><mo>+</mo></msubsup></math></span> and critical dimensionless particle diameter <span><math><msubsup><mi>d</mi><mrow><mi>p</mi><mn>50</mn></mrow><mo>+</mo></msubsup></math></span>. A semi-empirical aerosol resuspension model satisfying the S-Logistic function relationship is obtained and validated with multiple sets of experimental data, and showing good agreement between the model predictions and the experimental results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111652"},"PeriodicalIF":1.9,"publicationDate":"2025-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144313807","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":"SPARTA: A flux adjustment methodology to interpret complex experiments","authors":"Paul A. Ferney,&nbsp;Ben A. Baker,&nbsp;Mark D. DeHart","doi":"10.1016/j.anucene.2025.111623","DOIUrl":"10.1016/j.anucene.2025.111623","url":null,"abstract":"<div><div>To accurately determine reactivity from a detector count rate, correcting spatial effects is of prime importance. Simulation methodologies are often used for spatial correction, but they may introduce an additional source of uncertainty if the results of the experiment are also used as the input data for the simulation. This work presents a flux adjustment methodology that can infer experimental reactivity and correct spatial effects without the need for a simulation. It can process the detector signal(s) from a complex experiment, such as a heat balance measurement in the Transient Reactor Test Facility (TREAT) in which control rods are continuously adjusted to maintain a constant power. The methodology presented in this work successfully computed the reactivity and the local spatial variation of the flux from a generated signal. It also proved to be robust against noise and errors on kinetic parameters and provided a credible interpretation of a heat balance experiment in TREAT. An efficient flux adjustment method for complex experiments enables better experiment interpretation that is less reliant on nuclear data evaluations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111623"},"PeriodicalIF":1.9,"publicationDate":"2025-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144313810","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":"Radiological impact assessment of a hypothetical accident at the Zaporizhzhia nuclear power plant","authors":"Kyung-Suk Suh ,&nbsp;Sora Kim ,&nbsp;Kihyun Park ,&nbsp;Byung-Il Min ,&nbsp;Yoomi Choi ,&nbsp;Jiyoon Kim ,&nbsp;Min-Chae Kim ,&nbsp;Hyeonjeong Kim ,&nbsp;Kyeong-Ok Kim","doi":"10.1016/j.anucene.2025.111680","DOIUrl":"10.1016/j.anucene.2025.111680","url":null,"abstract":"<div><div>The atmospheric dispersion of radioactive materials and the resulting radiation dose were assessed for a hypothetical accident at the Zaporizhzhia Nuclear Power Plant in Ukraine. The release quantities of ¹³¹I and ¹³⁷Cs were assumed to be the same as those released into the atmosphere in the Chernobyl nuclear accident. The evaluation utilized atmospheric dispersion and dose assessment models, both of which are key components of the Radiological Accident Preparedness System in Korea (RAPS-K) developed by the Korea Atomic Energy Research Institute. Simulation results showed that radioactive plumes initially moved to the western across Europe, and later, some plumes were transported to the Asia due to westerly winds. Dose assessments revealed that effective radiation doses were showed above 1 mSv in certain areas near the Zaporizhzhia plant, while radiation exposure remained below 0.1 mSv across the rest of Europe, Asia, and North America. Especially, the thyroid dose due to ¹³¹I was presented about 19 mSv and 29 mSv, respectively in Kyiv and Odesa of Ukraine.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111680"},"PeriodicalIF":1.9,"publicationDate":"2025-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144321027","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":"Underlying mathematical relations for COMET calculations in problems with global symmetry","authors":"Dingkang Zhang,&nbsp;Farzad Rahnema","doi":"10.1016/j.anucene.2025.111639","DOIUrl":"10.1016/j.anucene.2025.111639","url":null,"abstract":"<div><div>This paper derives the underlying mathematical relations for reactor core problems with global reflection and/or rotation symmetry within the context of COMET’s incident flux response expansion theory. The derivation rigorously establishes the relationships between the expansion moments of the incoming and outgoing partial current across coarse mesh surfaces and their corresponding symmetric surfaces under various scenarios in the core. These global (i.e., whole-core) symmetry relations are then integrated into the hybrid stochastic deterministic coarse mesh transport code COMET, enabling the new code to model a portion of reactor cores which have reflection and/or rotation symmetry without increasing the number of unique coarse meshes in the precomputation of the COMET response function library. The new COMET code is numerically validated in Cartesian and Hexagonal geometries by using two sets of problems, namely, a set of stylized PWR benchmark core configuration with 1/8th reflection symmetry and a set of three Advanced High Temperature Reactor (AHTR) core configurations with 120° rotation symmetry. Results from these benchmark calculations demonstrate that the core eigenvalues and fission density distributions predicted by modeling only a portion of the core using the global symmetry relations are in excellent agreement with the full core results as expected. The difference in the core eigenvalues varies from 0 to 2 pcm for both the PWR and AHTR benchmark problems. The average relative differences in the fission density distributions range from 0.012% to 0.014% and from 0.044% to 0.060% for the PWR and AHTR problems, respectively. All the discrepancies fall within three standard deviations of the corresponding COMET uncertainties. Additionally, it is found that modeling just the symmetric portion of the cores speed up COMET by eight and three times for the PWR and AHTR core configurations, respectively.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111639"},"PeriodicalIF":1.9,"publicationDate":"2025-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144313808","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":"Physics Informed Neural Networks for the mixed dual form of the neutron diffusion equation with heterogeneous coefficients","authors":"Minh-Hieu Do,&nbsp;Karim Ammar,&nbsp;Nicolas Gerard Castaing,&nbsp;François Madiot","doi":"10.1016/j.anucene.2025.111607","DOIUrl":"10.1016/j.anucene.2025.111607","url":null,"abstract":"<div><div>Physics-Informed Neural Networks (PINNs), a popular deep learning framework for numerical approximations of Partial Differential Equations (PDEs), are investigated in this work to approximate the solution of the neutron diffusion equation, which is used in simulations of nuclear reactor cores. Moreover, this equation may have low regularity solution due to heterogeneous coefficients, which presents a challenge for the PINNs approach based on the primal form of the neutron diffusion equation. In this work, we study the PINNs approach for the mixed dual form of the neutron diffusion equation and aim to demonstrate that it can significantly improve the accuracy of the approximate solution, especially in cases with heterogeneous coefficients, compared to the primal approach. Besides, neural networks are typically based on the inverse power method for the k-eigenvalue problem, and it is well-known that this algorithm converges very slowly if the dominance ratio is high, as is commonly the case in several reactor physics applications. Therefore, we also discuss some acceleration methods for the PINNs approach applied to the k-eigenvalue problem. Several numerical test cases for the source and k-eigenvalue problems illustrate our purposes.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111607"},"PeriodicalIF":1.9,"publicationDate":"2025-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144313809","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":"CFD analysis on the safety performance of CSNS solid target under a postulated large loss of coolant accident","authors":"Zhiyu Yang ,&nbsp;Songbai Cheng ,&nbsp;Youlian Lu ,&nbsp;Songlin Wang ,&nbsp;Bin Zhou ,&nbsp;Tianjiao Liang ,&nbsp;Kai Wang ,&nbsp;Jianfei Tong","doi":"10.1016/j.anucene.2025.111654","DOIUrl":"10.1016/j.anucene.2025.111654","url":null,"abstract":"<div><div>This study examines the thermal behavior of the China Spallation Neutron Source (CSNS) Phase II target station during a hypothetical large loss of coolant accident (LOCA), where coolant channels are replaced by nitrogen gas at 0.11 MPa. Heat dissipation relies on natural convection and radiation. CFD simulations in STAR-CCM + analyze the impact of target material thermal conductivity, radiative heat transfer, and surface heat transfer coefficients on temperature distribution. A 20 % margin above the 500 kW baseline proton beam power and a 20 % reduction in target material thermal conductivity are applied. Without radiation modeling, the peak temperature reaches 490 ℃, but with the S2S radiation model, it drops to 420 ℃, highlighting radiation’s critical role. Increasing the surface heat transfer coefficient from 10 to 20 W/(m²·K) has a modest but notable effect.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111654"},"PeriodicalIF":1.9,"publicationDate":"2025-06-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144313676","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":"Plate-fin heat exchanger optimal design for industry using both single and multi-objective granularity based surrogate assisted Kho-Kho optimization","authors":"Subinaya Mohapatra ,&nbsp;Dushmanta Kumar Das ,&nbsp;Amit Kumar Singh","doi":"10.1016/j.anucene.2025.111605","DOIUrl":"10.1016/j.anucene.2025.111605","url":null,"abstract":"<div><div>All engineering devices stand on the basis of design. The mechanical design of Plate-fin heat exchanger (PFHE) always needs a better design to reduce cost while increasing productivity. For an optimal design of PFHE, three distinct issues: total heat transfer area, total volume, and total annual cost, are tackled using a novel meta-heuristic approach. This method is termed as the Granularity-based Surrogate-assisted Kho-Kho (GBSA-KKO) approach. The population is granulated into two subsets, i.e., fine grained and coarse grained and then different approximation methods are implemented with new infill criteria. Eight different benchmark functions are carried out to estimate the efficiency of the proposed GBSA-KKO approach. In conclusion, the proposed optimization approaches successfully minimize the objectives of total heat transfer area, total volume, and total annual cost, yielding values of 79.35 m<span><math><msup><mrow></mrow><mrow><mn>2</mn></mrow></msup></math></span>, 0.00247 m<span><math><msup><mrow></mrow><mrow><mn>3</mn></mrow></msup></math></span>, and 838.218$, respectively.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"223 ","pages":"Article 111605"},"PeriodicalIF":1.9,"publicationDate":"2025-06-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144307011","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}