Annals of Nuclear Energy最新文献

{"title":"Analysing hydrogen behaviour in liner claddings using OFFBEAT fuel performance code","authors":"L. Verma ,&nbsp;I. Clifford ,&nbsp;P. Konarski ,&nbsp;A. Scolaro ,&nbsp;H. Ferroukhi","doi":"10.1016/j.anucene.2025.111559","DOIUrl":"10.1016/j.anucene.2025.111559","url":null,"abstract":"<div><div>Hydrogen generated during the waterside corrosion of zirconium claddings can diffuse within the material, forming hydrides that alter the mechanical properties of the claddings. These hydrides may lead to detrimental effects such as delayed hydride cracking and embrittlement, compromising the integrity of the claddings. Liner claddings, which have been particularly used in Swiss reactors, have been shown to aid in minimizing these detrimental effects of hydrogen behaviour on zirconium claddings. In this paper, hydrogen behaviour models for liner claddings have been incorporated into the multi-dimensional, thermo-mechanical fuel performance code, OFFBEAT. Numerical analyses using OFFBEAT are then compared to out-of-pile experimental data on two liner cladding types commonly used in Swiss reactors. The findings indicate that OFFBEAT’s predictions closely align with experimental observations, demonstrating hydrogen migration toward the substrate-liner interface (SLI). This research confirms OFFBEAT’s capability to simulate hydrogen behavior in liner claddings, supporting further exploration of alternative cladding configurations moving forward towards more advanced technological fuels for nuclear safety.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111559"},"PeriodicalIF":1.9,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138186","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":"Direct NeTS sampling of nuclear graphite S(α,β,T) in Serpent","authors":"Jonathan P.W. Crozier,&nbsp;Ayman I. Hawari","doi":"10.1016/j.anucene.2025.111549","DOIUrl":"10.1016/j.anucene.2025.111549","url":null,"abstract":"<div><div>For advanced reactor applications, Neural Thermal Scattering (NeTS) modules were developed to predict the thermal scattering law (TSL or <span><math><mrow><mi>S</mi><mo>(</mo><mi>α</mi><mo>,</mo><mi>β</mi><mo>,</mo><mi>T</mi><mo>)</mo></mrow></math></span>) of a nuclear graphite neutron moderator. NeTS are multi-layer, feedforward artificial neural networks, which act as universal function approximators designed for TSL datasets. In this case, a 4-layer neural network with 164 neurons per layer is trained using <em>FLASSH</em> evaluated data in PyTorch and serialized as a <em>torchscript</em> dictionary to predict <span><math><mrow><mi>S</mi><mo>(</mo><mi>α</mi><mo>,</mo><mi>β</mi><mo>,</mo><mi>T</mi><mo>)</mo></mrow></math></span> on-the-fly. Relative, absolute and maximum percent deviations of NeTS from File 7 data generated using the <em>FLASSH</em> code are on the order of 0.01%, 0.1% and 1%, respectively, with low inference latencies of 0.000172 s per <span><math><mrow><mi>S</mi><mfenced><mrow><mi>α</mi><mo>,</mo><mi>β</mi><mo>,</mo><mi>T</mi></mrow></mfenced></mrow></math></span> at a given temperature. Capturing the full dimensionality of possible inelastic neutron-lattice interactions, NeTS functionality is embedded in the Serpent Monte Carlo code, where <span><math><mrow><mi>S</mi><msub><mfenced><mrow><mi>α</mi><mo>,</mo><mi>β</mi><mo>,</mo><mi>T</mi></mrow></mfenced><mrow><mi>NeTS</mi></mrow></msub></mrow></math></span> sampling is conducted on-the-fly and compared to ACE look-up-tables for predicting TREAT criticality. k-eff differences between sampling algorithms of 6 pcm are observed and are within the order of Monte Carlo uncertainty. Compared to discrete and continuous-energy ACE files (30 MB and 131 MB per temperature), the NeTS format is on the order of 200–300 kB for a continuous-temperature, interpolation-free representation of <span><math><mrow><mi>S</mi><mo>(</mo><mi>α</mi><mo>,</mo><mi>β</mi><mo>,</mo><mi>T</mi><mo>)</mo></mrow></math></span> and cross sections. NeTS-in-Serpent runtimes comparable with ACE look-up tables are achieved by scaling NeTS for high performance computing architectures with hybrid OpenMP + MPI parallelization. This work validates a novel, self-contained reactor physics framework for predictive cross sections, and demonstrates a general methodology for embedding modern machine learning libraries within existing neutronic analysis frameworks.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111549"},"PeriodicalIF":1.9,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138187","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":"Effect of buoyancy-induced natural convection heat transfer on thermal qualification of lead-shielded radioactive material transport packages","authors":"Sampath Bharadwaj Kota,&nbsp;Seik Mansoor Ali","doi":"10.1016/j.anucene.2025.111570","DOIUrl":"10.1016/j.anucene.2025.111570","url":null,"abstract":"<div><div>Radioactive materials (RM) are stored and transported in well-shielded packages that require certification by the nuclear regulatory authority. Certification is subject to successful demonstration of compliance to a series of tests. Among these, the thermal test is intended to simulate a 30-minute, fully engulfing fire, that may occur during a transport accident involving spillage of large quantities of hydrocarbon fuel. In the present work, the thermal response of a Type-B(U) transport package to fire exposure is investigated by applying CFD techniques. The phase-change processes are captured in detail and the effect of buoyancy-induced natural convection in enhancing the heat transfer and melt fraction is brought out. The duration of convection dominated heat transfer to the package is clearly demarcated from the conduction dominated heat transfer by identifying a cross-over point. The instantaneous heat transfer rates into the package are analysed to bring out the fraction of heat stored as sensible or latent heat. The effect of specifying fire boundary condition using HC-air curve, ISO-834 curve or a constant ambient temperature of 800 °C is also highlighted by means of parametric studies. Finally, some comments on the effect of RM package size on the heat transfer aspects is provided.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111570"},"PeriodicalIF":1.9,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144154425","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Study on non-uniform control rod with BeO reflectors","authors":"Feilong Liu ,&nbsp;Xiaoqin Nie ,&nbsp;Du Liu ,&nbsp;Xiaoan Li ,&nbsp;Cheng Liu ,&nbsp;Zhengguo Chen","doi":"10.1016/j.anucene.2025.111589","DOIUrl":"10.1016/j.anucene.2025.111589","url":null,"abstract":"<div><div>When the non-uniform control rods are used in power control system of Pressurized Water Reactor(PWR), its outer neutron absorbers will absorb neutrons throughout its life, which will reduce the fuel burnup level. To reduce this influence, outer absorbers of non-uniform control rod are replaced by the BeO reflectors. Without considering the external reflectors and the arrangement of combustible poison rods, The N = 2 type AB non-uniform control rod system of Qinshan nuclear power plant in China is established by MCNP5. And then the core axial neutron flux densities are calculated when non-uniform control rod is at five thickness BeO reflectors. The simulation results show that the non-uniform control rod with BeO reflectors can not only keep control functions of axial power deviation and power level control as without reflector, but also increase the core neutron flux density by increasing the thickness of the BeO reflectors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111589"},"PeriodicalIF":1.9,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138182","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":"Investigation of thermo-mechanical behaviour of 220MWe IPHWR channel under accdiental scenario (LOCA)","authors":"Shivam Raj ,&nbsp;Apurb Ayush Topno ,&nbsp;Kumar Piyush Aman ,&nbsp;Mushtaque Momin ,&nbsp;Subodh Kumar Yadav ,&nbsp;Mukesh Sharma","doi":"10.1016/j.anucene.2025.111604","DOIUrl":"10.1016/j.anucene.2025.111604","url":null,"abstract":"<div><div>This study focuses on Indian Pressurized Heavy Water Reactor (IPHWR) channel under Loss of Coolant Accident (LOCA) and Emergency Core Cooling System<!--> <!-->failure. The study utilizes numerical analysis based on experimental data to investigate temperature distribution, structural deformation, and heat transfer between the pressure tube (PT) and calandria tube (CT). As observed in past studies, the excessive heat from the fuel bundle released after the aforementioned accident causes sagging, further leading to PT-CT contact, which acts as a crucial heat sink, essentially acting as a passive countermeasure against meltdown. The results match with the experimental data and key findings indicates significant temperature variations in both PT and CT during PT-CT contact, which also mitigates significant amounts of heat. Further as the temperature increases the risk of PT melting persists at high amounts of heat.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111604"},"PeriodicalIF":1.9,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138189","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":"Verification of practical modeling approaches of an MSLB event simulation: A CFD contribution","authors":"Ezequiel Fogliatto,&nbsp;Ivor Clifford,&nbsp;Alexander Vasiliev,&nbsp;Abdelhamid Dokhane,&nbsp;Ferroukhi Hakim","doi":"10.1016/j.anucene.2025.111572","DOIUrl":"10.1016/j.anucene.2025.111572","url":null,"abstract":"<div><div>This article describes the development and application of a Computational Fluid Dynamics (CFD) model to simulate the fluid flow, heat transfer, and boron distribution during a Main Steam Line Break (MSLB) event in a three-loop pre-Konvoi Nuclear Power Plant. The CFD simulations were used to produce time-varying flow maps at the core inlet, which were integrated into a transient SIMULATE-3 K core model using a loose-coupling methodology. The simulations predict complex flow patterns within the reactor pressure vessel, characterized by pronounced temperature stratification in the downcomer and a well-defined cold plume descending directly from the affected cold leg. This coolant behavior significantly influences flow and boron distributions, with notable deviations at the core inlet compared to traditional static mixing assumptions. Transient reactivity calculations conducted with SIMULATE-3 K, utilizing mixing maps derived from the CFD simulations and an in-house mixing tool, show a larger margin to recriticality than traditional static homogeneous mixing maps.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111572"},"PeriodicalIF":1.9,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138188","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 new format for the elastic scattering differential cross section covariance of charged particles","authors":"Xu Han ,&nbsp;Jie Liu ,&nbsp;Zhen-Peng Chen ,&nbsp;Tao Ye","doi":"10.1016/j.anucene.2025.111556","DOIUrl":"10.1016/j.anucene.2025.111556","url":null,"abstract":"<div><div>For the elastic scattering of charged particles, there is no clear format in the ENDF-6 document to publish the covariance data of the differential cross sections. Since the covariance data is a function of both energy and angle, the publication of the raw data will make the data file extremely large and is difficult to be used in the nuclear data processing programs. In this paper, we proposed a new format for publishing this data. The angular distributions of correlation coefficients are fitted using a superposition equation and the fitting parameter values along with the standard deviation data are published, which can be used to reconstruct the covariance matrix. Based on the Generalized Reduced R-matrix theory, the RAC program (R-matrix analysis code) was employed to evaluate the experimental data of all nuclear reaction channels associated with the <!--> ⁷Be system. The <!--> ⁶Li(p, p)⁶Li reaction of the <!--> ⁷Be system were selected as an example for testing the proposed data format. The calculated results show that the amount of released data with the new format is only 1/10 of the original covariance raw data and the accuracy is reduced by only 5.21%, which indicates the proposed new format is effective and reliable.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111556"},"PeriodicalIF":1.9,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144134987","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 of Cl-35 nuclear data on a Molten Chloride Fast Reactor","authors":"Jean-Baptiste Valentin ,&nbsp;Massimiliano Fratoni ,&nbsp;Daniel Siefman ,&nbsp;Ludovic Jantzen ,&nbsp;Mathieu Hursin","doi":"10.1016/j.anucene.2025.111529","DOIUrl":"10.1016/j.anucene.2025.111529","url":null,"abstract":"<div><div>This study investigates the impact of ³⁵Cl nuclear data uncertainties on the neutronics of Molten Chloride Fast Reactors (MCFR), specifically focusing on two models: MCFR-C and MCFR-D. Using the Monte Carlo code SERPENT2, a comprehensive sensitivity analysis and uncertainty quantification was conducted for both initial and equilibrium fuel compositions. This study was done synchronously with nuclear data evaluators at Los Alamos National Laboratory who were creating a new evaluation for ³⁵Cl. These findings reveal that the new ³⁵Cl evaluation has minimal effect on core neutronics for these designs (however could have a significant impact for a different flux spectrums), but significantly reduces the uncertainty in the effective neutron multiplication factor (<span><math><msub><mrow><mi>k</mi></mrow><mrow><mtext>eff</mtext></mrow></msub></math></span>) to <span><math><mo>∼</mo></math></span>1000 pcm (from <span><math><mo>∼</mo></math></span>1400 pcm). A robust method and workflow was developed to propagate uncertainties through SERPENT’s sensitivity analysis, incorporating the Monte Carlo statistical uncertainties and using a Positive Semi-Definite correction algorithm for nuclear covariance data. Though limited by significant computational time and memory usage, this approach offers a reliable method for uncertainty propagation offering valuable insights into the static and uncertainty parameters of MCFR-C and MCFR-D reactors, thereby contributing to the advancement of MCFR technology. It also provides a valuable example of how downstream applied neutronics can work together with nuclear data evaluators to improve reactor analyses.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111529"},"PeriodicalIF":1.9,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144134988","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":"Reactor power regulation of a small pressurized water reactor based on linear active disturbance rejection control with gain scheduling","authors":"Ziqi Fan,&nbsp;Peiwei Sun,&nbsp;Xinyu Wei","doi":"10.1016/j.anucene.2025.111585","DOIUrl":"10.1016/j.anucene.2025.111585","url":null,"abstract":"<div><div>The small pressurized water reactor (SPWR) adopts the integral design, which has the advantages of simple structure, small volume and high safety. However, SPWR is a system with nonlinear and time-varying complex dynamics. Under large and abrupt load regulation, the traditional power control system has the problem of large overshoot. To solve this problem, a linear active disturbance rejection reactor control with gain scheduling is proposed. In this control scheme, the power deviation is handled by linear active disturbance rejection controller and the coolant temperature deviation is passed by PI controller, which jointly adjust the control rod. Thus, the power can quickly adapt to the load change while the temperature deviation can meet the control requirements. The controller parameters optimized by genetic algorithm under different power levels are expressed as a function by linear interpolation method to form an offline database. The optimal parameter of the control system is obtained by querying the offline database in real time. Simulation tests are carried out in the typical conditions of forced circulation and natural circulation. The results show that under the designed control system the power overshoot and settling time can be effectively reduced compared with the traditional control system.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111585"},"PeriodicalIF":1.9,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138185","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":"New approach to evaluate axial deformation of PWR fuel assembly on criticality safety of transport packages under impact accidents","authors":"Xinling Dai ,&nbsp;Yanmin Zhang ,&nbsp;Junxiang Xu ,&nbsp;Dechang Cai ,&nbsp;Jin Cai","doi":"10.1016/j.anucene.2025.111558","DOIUrl":"10.1016/j.anucene.2025.111558","url":null,"abstract":"<div><div>In accordance with pertinent regulations, Pressurized Water Reactor (PWR) fuel assemblies transport packages must maintain subcriticality across all transport scenarios, including those involving accidents. Accidents can induce buckling in fuel rods, thereby altering the local pin pitch. An expanded pitch can adversely affect criticality safety; hence, it is customary to presume maximum lattice expansion across the entire active zone to anticipate the worst potential accidents. However, mechanical analysis reveal that lattice expansion is confined to the impact end. Consequently, the original evaluative model overestimated <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>e</mi><mi>f</mi><mi>f</mi></mrow></msub></math></span> without necessity. This study refines the evaluative model for PWR fuel assemblies under transportation accidents by concerning axial deformation, thereby releasing the safety margin. The refined model can be seamlessly integrated with non-uniform arrangement of fuel rods. These advantages make it more suitable as an envelope model for criticality safety analysis of fuel transport packages.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"222 ","pages":"Article 111558"},"PeriodicalIF":1.9,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138183","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}