Annals of Nuclear Energy最新文献

{"title":"Research reactors: Global perspectives and insights into Saudi Arabia’s advancements and future prospects","authors":"Afaque Shams ,&nbsp;Ouadie Kabach ,&nbsp;Husam Khalefih ,&nbsp;Khaled Al-Athel ,&nbsp;Tomasz Kwiatkowski ,&nbsp;Gaweł Madejowski","doi":"10.1016/j.anucene.2025.111740","DOIUrl":"10.1016/j.anucene.2025.111740","url":null,"abstract":"<div><div>Research reactors play a crucial role in harnessing nuclear energy for diverse applications across various fields, serving as critical infrastructure for science, technology, and health sectors worldwide. This paper provides a comprehensive review of the current status, utilization, and design aspects of research reactors worldwide, with the goal of informing Saudi Arabia’s emerging research reactor program. It explores their classification based on criticality status and power levels, highlighting their flexibility compared to standard Light Water Reactors (LWRs). Unlike LWRs, research reactors employ diverse fuel materials, often enriched to levels exceeding 5 wt%. Key applications, including radioisotope production, neutron activation analysis, and beam port applications, are discussed in detail. Special attention is given to the production of critical isotopes such as I-131, Te-99, and Ir-192, emphasizing the global challenges posed by shortages and operational constraints among major producers. The paper also reviews the status of Saudi Arabia’s research reactor program, shedding light on the sectors involved in the project’s development and the potential contributions to the nuclear energy landscape. This includes an assessment of local needs for isotope production, education, and industrial testing, particularly in the eastern province. The importance of expanding isotope production capacity to meet growing demands is underscored, along with strategies to overcome existing challenges. This study provides insights into the evolving role of research reactors and underscores their significance in advancing scientific research, medical advancements, and industrial applications. It also highlights the strategic opportunity for Saudi Arabia to address regional gaps and build nuclear capacity by leveraging lessons from international research reactor programs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111740"},"PeriodicalIF":1.9,"publicationDate":"2025-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654998","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":"Neutron transmission experiment with PTFE at AKR-2","authors":"Michal Košťál ,&nbsp;Marek Zmeškal ,&nbsp;Tomáš Czakoj ,&nbsp;Juan José Gómez Rodriguez ,&nbsp;Marco Viebach ,&nbsp;Alexander Knospe ,&nbsp;Vincent Melzer ,&nbsp;Sascha Weichel ,&nbsp;Carsten Lange ,&nbsp;Antonio Hurtado ,&nbsp;Evžen Losa ,&nbsp;Jan Šimon ,&nbsp;Stanislav Simakov ,&nbsp;Martin Schulc ,&nbsp;Evžen Novák ,&nbsp;Tomáš Peltan ,&nbsp;Zdeněk Matěj ,&nbsp;Roberto Capote","doi":"10.1016/j.anucene.2025.111741","DOIUrl":"10.1016/j.anucene.2025.111741","url":null,"abstract":"<div><div>Fluorine is a crucial element for the nuclear industry and technology due to its application in nuclear fuel production as uranium hexafluoride (UF₆), its use of Teflon® in criticality experiments, and its use in coolant/fuel materials in several Molten Salt Reactor (MSR) concepts currently being designed. There have been many efforts to improve the neutron reaction cross sections for fluorine. The neutron broad-beam transmission experiment through the PTFE (PolyTetraFluoroEthylene) block at the Dresden AKR-2 reactor has been performed to validate existing evaluated cross-section libraries from 1 to 10 MeV of incident neutron energies. Calculations with studied data sets (ENDF/B-VIII.0, INDEN, JEFF-3.3, and JENDL-5) show satisfactory agreement with the experiment in the energy region above 2.1 MeV within the uncertainty range. The measured neutron transmission shows a<!--> <!-->systematic disagreement in a calculation to experiment comparison in the energy region 1–2.1 <!--> <!-->MeV. This behavior is consistent with the measured ²⁵²Cf(s.f.) neutron leakage in the PTFE block with dimensions of 60 × 50 × 50 cm. Improvement of evaluated cross sections below 2 MeV of neutron incident energy is needed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111741"},"PeriodicalIF":1.9,"publicationDate":"2025-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144633741","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 numerical stabilization model for the half-boundary method in solving heat conduction problems","authors":"Haoyu Yin ,&nbsp;Xiaojing Liu ,&nbsp;Xiang Chai ,&nbsp;Hui He ,&nbsp;Lianjie Wang ,&nbsp;Bin Zhang ,&nbsp;Tengfei Zhang","doi":"10.1016/j.anucene.2025.111734","DOIUrl":"10.1016/j.anucene.2025.111734","url":null,"abstract":"<div><div>The half-boundary method (HBM), an extension of the boundary element method, demonstrates superior accuracy and computational efficiency in nuclear fuel rod heat transfer simulations. However, its numerical stability under inappropriate time steps and iteration schemes remains unresolved, a problem systematically addressed in this work for the first time. Von Neumann analysis reveals that temporal stability requires the implicit weighting parameter <span><math><mrow><mi>θ</mi></mrow></math></span> of time discretization scheme to exceed 0.5. Spectral radius analysis for spatial iteration indicates stability in steady-state but instability in transient-state conditions. To address spatial instability, we analyze error amplification mechanisms and propose a stabilization model that maps acceptable error thresholds to time step sizes, thereby ensuring stable spatial iterations. Our analysis further demonstrates that computational stability improves as the mesh Fourier number deviates from unstable ranges. The proposed model is successfully validated through application to realistic nuclear fuel rod heat conduction problems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111734"},"PeriodicalIF":1.9,"publicationDate":"2025-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654991","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 of the effect of burnup depth on U3Si2-Al dispersion helical cruciform fuel","authors":"Biao Li ,&nbsp;Wenchao Su ,&nbsp;Shilei Li ,&nbsp;Qiang Fu ,&nbsp;Shengli Mao ,&nbsp;Yong Shuai","doi":"10.1016/j.anucene.2025.111725","DOIUrl":"10.1016/j.anucene.2025.111725","url":null,"abstract":"<div><div>The U₃Si₂-Al dispersed helical cruciform fuel (HCF) element, distinguished by high burnup tolerance and superior thermal conductivity, represents an innovative nuclear fuel design. This study employs finite element analysis (FEA) to investigate the thermo-mechanical behaviors of HCF element under varying U₃Si₂ particle loadings and burnup levels. The results indicate that, at 5 % burnup, increasing the fuel particle volume fraction from 10 % to 30 % raises the maximum fuel temperature from 593.51 K to 598.05 K, while stress concentrates at the concave arc, where the matrix contacts the cladding, reaching a peak stress of 344.95 MPa. At 20 % fuel particle volume fraction under irradiation, burnup progression from 5 % to 20 % increases maximum stress by 46.55 MPa, while temperature distribution remains unaffected. The analysis reveals a transition in stress-dominant factors: at low burnup, the U₃Si₂ particle volume fraction governs stress distribution, whereas at high burnup, irradiation-induced swelling becomes the primary driver of deformation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111725"},"PeriodicalIF":1.9,"publicationDate":"2025-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144633740","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 quasi-differential measurement of neutron scattering from tantalum between 0.65 MeV and 20 MeV","authors":"Gregory J. Siemers ,&nbsp;Yaron Danon ,&nbsp;Sukhjinder Singh ,&nbsp;Katelyn C. Keparutis ,&nbsp;Peter J. Brain ,&nbsp;Benjamin H. Wang ,&nbsp;Adam M. Daskalakis ,&nbsp;Devin P. Barry ,&nbsp;Michael J. Rapp ,&nbsp;Robert C. Block","doi":"10.1016/j.anucene.2025.111676","DOIUrl":"10.1016/j.anucene.2025.111676","url":null,"abstract":"<div><div>Quasi-differential neutron scattering measurements of elemental tantalum (Ta) and carbon were performed at the Gaerttner Linear Accelerator Center at Rensselaer Polytechnic Institute using neutron time-of-flight spectroscopy. An array of eight EJ-301 organic liquid scintillator detectors measured neutron interactions between the pulsed white neutron source and the samples of interest. Signals were converted to digital pulses using a new 10-bit Struck SIS-3305 digitizer, which enhanced neutron detection efficiency above 3 MeV over previous measurements. An improved pulse shape analysis technique was developed and used to discriminate between incident neutron and photon interactions with the scintillator. Neutron time-of-flight histograms were compared with detailed Monte Carlo radiation transport simulations using MCNP6.3. The elemental carbon measurement was reproduced well by the ENDF/B-VIII.0 carbon evaluation, thus verifying the experimental methods and simulation geometry, and providing a normalization to the experimental data. The average deviation between the carbon measurement and evaluation, 4.0%, was adopted as the systematic uncertainty of the experiment. Evaluated neutron scattering and emission data of <span><math><msup><mrow></mrow><mrow><mn>181</mn></mrow></msup></math></span>Ta from the ENDF/B-VIII.0, JEFF-3.3, ENDF/B-VIII.1, JENDL-5.0, and TENDL-2023 (updated) libraries were compared with the Ta measurement. The ENDF/B-VIII.0, JEFF-3.3, and TENDL-2023u <span><math><msup><mrow></mrow><mrow><mn>181</mn></mrow></msup></math></span>Ta evaluations significantly overpredict Ta neutron scattering between 0.75 MeV and 2.5 MeV at backward scattering angles. In contrast, the ENDF/B-VIII.1 <span><math><msup><mrow></mrow><mrow><mn>181</mn></mrow></msup></math></span>Ta evaluation slightly underpredicts the experimental data in this energy range at all measured angles. Generally good agreement is observed between all evaluations and the experimental data above 5 MeV at all angles. The JENDL-5.0 and ENDF/B-VIII.1 <span><math><msup><mrow></mrow><mrow><mn>181</mn></mrow></msup></math></span>Ta evaluations show the best agreement with the experimental findings.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111676"},"PeriodicalIF":1.9,"publicationDate":"2025-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144655049","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":"FissionIST: A hardware-based research reactor simulator","authors":"E.D. Ferreira ,&nbsp;L.M. Cabral ,&nbsp;M. Felizardo,&nbsp;P. Lima,&nbsp;J.G. Marques","doi":"10.1016/j.anucene.2025.111750","DOIUrl":"10.1016/j.anucene.2025.111750","url":null,"abstract":"<div><div>FissionIST is a hardware-based simulator of the 1 MW Portuguese Research Reactor (RPI) at <em>Instituto Superior Técnico</em> using a control console of the reactor. It was specifically designed to support education and training programs after the permanent shutdown of the RPI. The input from neutron detectors is simulated using programmable pulse and current generators, taking into consideration the assumed position of the control rods and other parameters. FissionIST simulates the time behavior of reactor power and reactivity by using 6-group point kinetics equations. A Windows-based workstation runs the application software entirely developed in Python. In this work, the system is described, and tests involving the insertion of both positive and negative reactivity are presented, with the results compared to data previously acquired at the RPI.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111750"},"PeriodicalIF":1.9,"publicationDate":"2025-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144655048","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":"Studies of the influence of heat pipe model and coupling strategy in heat pipe microreactor simulations using Sockeye","authors":"Joshua E. Hansel,&nbsp;Javier Ortensi,&nbsp;Mustafa K. Jaradat","doi":"10.1016/j.anucene.2025.111694","DOIUrl":"10.1016/j.anucene.2025.111694","url":null,"abstract":"<div><div>This work demonstrates various modeling approaches for thermally coupling heat pipes to a graphite monolithic block in a microreactor context, accounting for the presence of a gap. We show that heat pipe model selection has a significant impact on the transient behavior of a three-dimensional coupled microreactor assembly problem by comparing results obtained using different heat pipe models from the heat pipe code Sockeye. Additionally, we show that the choice of coupling strategy has significant consequences for both accuracy and performance. We found that our heat flux transfer strategy exhibited greater accuracy than our temperature transfer strategy, even when comparing a loose coupling of the heat flux transfer strategy to a tight coupling of the temperature transfer strategy. Although this research uses only theoretical test problems, it provides important insights in modeling thermal fluids phenomena in heat pipe microreactors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111694"},"PeriodicalIF":1.9,"publicationDate":"2025-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144654999","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":"KARATE-1200: An enhanced neutron transport code for VVER-1200 reactor physics calculations","authors":"Gy. Hegyi,&nbsp;E. Temesvári","doi":"10.1016/j.anucene.2025.111735","DOIUrl":"10.1016/j.anucene.2025.111735","url":null,"abstract":"<div><div>Comprehensive safety analysis is crucial for the deployment of new reactor designs. It requires accurate prediction of reactivity caused by temperature, boron concentration changes or control rod movement, as well as the 3D power, temperature, and burnup distributions during the cycle. Moreover, the reactor’s behavior under accident conditions has to be assessed. This necessitates advanced neutron transport codes capable of detailed modeling of heterogeneous core structures. It uses detailed meshing to model inhomogeneous structures, following the latest VVER core, and enables accurate prediction of reactor parameters during the burnup. One possible solution is to develop further a neutron code that has already been proven in practice for similar tasks.</div><div>This article presents the capabilities and performance of KARATE-1200, an in-house developed deterministic neutron transport code designed for third-generation VVER reactors. Building upon the 40-year legacy of the KARATE code used at VVER-440 NPPs, KARATE-1200 incorporates significant enhancements to improve VVER modeling from pin-cell to coarse-mesh levels, achieving good agreement with reference data. The KARATE-1200 code package, incorporating MULTICELL for group constant generation and GLOBUSKA-1200 for criticality calculations, has been verified against benchmark solutions and validated against published measurements. Safety-related parameters of the VVER-1200 core, based on data from the Novovoronezh II NPP, were calculated. Simulated reactivity coefficients using KARATE-1200 show agreement within ± 3 % of published measurements, depending on the coefficient type. Furthermore, key safety parameters, such as the isothermal re-criticality temperature, also demonstrate good agreement with literature values.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111735"},"PeriodicalIF":1.9,"publicationDate":"2025-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144614030","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":"Complementary test of three-dimensional TROSE experiments for China advanced passive PWR under IVR strategy","authors":"Kun Han ,&nbsp;Lian Chen ,&nbsp;Zongyang Li ,&nbsp;Huajian Chang ,&nbsp;Fangfang Fang","doi":"10.1016/j.anucene.2025.111726","DOIUrl":"10.1016/j.anucene.2025.111726","url":null,"abstract":"<div><div>Understanding the coupling heat transfer of the three-layer molten pool in the lower head aids in assessing the thermal load on the reactor pressure vessel wall and its integrity. Following 3 previous tests conducted on the TROSE experimental setup, the Test-T4 condition was carried out, featuring approximately equal internal heating power density distributions in the oxide and heavy metallic layers. The simulants employed for the three layers were mineral oil, water, and Cerrobend alloy, with thicknesses of 0.2 m, 0.7 m, and 0.3 m, respectively. The experimental results indicate that the normalized sideward heat flux values ranged from 0.32 to 1.58, with minimum and maximum values located at the bottom and top of the oxide layer. Shell formation in the heavy metallic layer across all tests results in a dominant conduction mode, leading to a decreasing distribution trend with increasing polar angle, contrary to various correlations’ predictions.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111726"},"PeriodicalIF":1.9,"publicationDate":"2025-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144614028","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":"Simulation and control system design of AP1000 based on APROS","authors":"Tianqi Luan ,&nbsp;Ting Yang","doi":"10.1016/j.anucene.2025.111689","DOIUrl":"10.1016/j.anucene.2025.111689","url":null,"abstract":"<div><div>The coordinated control of nuclear power plants aims to maintain the stable operation of primary and secondary loops by regulating key parameters such as reactor power and axial offset, coolant temperature and pressure, steam generator water level, and steam flow rate and pressure, ensuring operational safety and system stability. In this study, a one-dimensional neutronics and thermal-hydraulic coupled core model of the AP1000 was established based on the APROS platform, and a coordinated control strategy was developed to regulate key parameters under typical operations. Simulations under steady-state and transient conditions were performed, and the control performance was compared against benchmark data. Results demonstrate that the proposed control system offers excellent load-following and coordination capabilities. Additionally, by integrating the AP1000-specific axial offset control strategy, the reactor power control system was optimized. A 10% step-load reduction transient simulation shows that the strategy maintains AO within the control band while improving regulation precision and dynamic response. The proposed modeling and control methods provide an effective approach for AP1000 simulation and control system design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"224 ","pages":"Article 111689"},"PeriodicalIF":1.9,"publicationDate":"2025-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144614027","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}