Annals of Nuclear Energy最新文献

{"title":"A hybrid L1-Chebyshev-HPM approach for solving fractional neutron diffusion equations with delayed neutrons","authors":"Ujwal Warbhe","doi":"10.1016/j.anucene.2025.111729","DOIUrl":"10.1016/j.anucene.2025.111729","url":null,"abstract":"<div><div>This work presents a novel derivation of a fractional neutron diffusion equation that incorporates memory effects and anomalous transport phenomena observed in nuclear reactor cores. Starting from the classical neutron diffusion model, the derivation extends the framework through the introduction of fractional calculus in the constitutive relations. The resulting formulation accommodates subdiffusive behavior and provides a more accurate description of delayed neutron kinetics. To numerically solve the derived fractional equation, a hybrid algorithm is developed that integrates an L1 finite difference approximation for temporal discretization, Chebyshev spectral collocation for spatial accuracy, and the Homotopy Perturbation Method (HPM) to treat nonlinearity. The method is rigorously analyzed for convergence and stability, yielding exponential spatial convergence and near-first-order temporal accuracy. Comprehensive numerical experiments, including step, ramp, and sinusoidal reactivity cases, demonstrate the superior accuracy and computational efficiency of the proposed scheme compared with traditional methods (e.g., finite element and B-spline collocation (Roul et al., 2020)). Unlike prior works, this framework uniquely combines spectral accuracy with fractional calculus, addressing gaps in modeling anomalous neutron transport in heterogeneous media (Espinosa-Paredes, 2023). This new framework offers a robust tool for reactor dynamics analysis under anomalous diffusion regimes, with potential applications in reactor safety and control by enabling high-fidelity simulations of memory-dependent transport.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111729"},"PeriodicalIF":2.3,"publicationDate":"2025-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750694","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":"Influence of geyser boiling phenomenon on the transient operating characteristics of solid-state reactor","authors":"Bo Dong ,&nbsp;Suyi Zhang ,&nbsp;Yugao Ma ,&nbsp;Kun Cheng ,&nbsp;Zhifang Qiu ,&nbsp;Feng Li ,&nbsp;Xiaoqiang He","doi":"10.1016/j.anucene.2025.111768","DOIUrl":"10.1016/j.anucene.2025.111768","url":null,"abstract":"<div><div>As the key component of heat removal inside the reactor, heat pipes can induce the heat-transfer oscillations of the reactor due to geyser boiling of alkali metal inside it. The geyser boiling exacerbates thermal fatigue damage to core components and heat pipes within the solid-state core. In this paper, a temperature oscillation function of heat pipe is imposed in Mega-Power solid-state reactor by using the heat pipe reactor transient analysis code (HPRTRAN). The geyser boiling inside heat pipes is simulated and the effects of heat pipe parameters and oscillation characteristics on the transient operation characteristics are further investigated. The results show that when the fill ratio increases from 0.25 to 0.75, the mean value of the core power oscillation deviates from its original value with the deviation magnitude reaching 20 % of full power. For the reactor startup process, geyser boiling results in amplitude of peak stress of fuel up to 3 MPa at 50 % FP.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111768"},"PeriodicalIF":2.3,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750690","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":"Numerical investigation of mixed and natural convection in 19-pin fuel assemblies with wire-wrapped based on the SST k-ω-DHFM model","authors":"Yuefeng Guo ,&nbsp;Xingkang Su ,&nbsp;Guan Wang ,&nbsp;Huan Lin ,&nbsp;Cong Lin ,&nbsp;Long Gu","doi":"10.1016/j.anucene.2025.111761","DOIUrl":"10.1016/j.anucene.2025.111761","url":null,"abstract":"&lt;div&gt;&lt;div&gt;In Lead-cooled Fast Reactors (LFRs), liquid metals such as lead or lead–bismuth eutectic (LBE) are used as coolants. The coolant circulation occurs either through forced convection or natural circulation. Natural circulation is driven by density variations induced by temperature differences, enabling heat transfer without mechanical pump. In natural circulation, buoyancy dominates the flow, which may be classified as natural or mixed convection based on the interplay between inertial and buoyant forces. Due to the high thermal diffusivity of liquid metals, the temperature boundary layer is significantly thicker than the velocity boundary layer, leading to the failure of the Reynolds analogy method based on the Simple Gradient Diffusion Hypothesis (SGDH). This discrepancy poses a challenge for accurately closing the turbulent heat flux transport equation, thereby hindering reliable prediction of the temperature field. To address this challenge, a second-order differential heat flux model (DHFM) has been developed based on OpenFOAM, which is a second-order five-equation heat flux model for liquid metals under buoyancy effects. By coupling the SST &lt;em&gt;k&lt;/em&gt;-&lt;em&gt;ω&lt;/em&gt; turbulence model with the DHFM model, an SST &lt;em&gt;k&lt;/em&gt;-&lt;em&gt;ω&lt;/em&gt;-DHFM model is established, specifically designed for analyzing mixed and natural convection of liquid metals. This study takes the natural circulation experiment of NACIE-UP as a benchmark to evaluate the applicability of the SST &lt;em&gt;k&lt;/em&gt;-&lt;em&gt;ω&lt;/em&gt;-DHFM model. Results indicated that the SST &lt;em&gt;k&lt;/em&gt;-&lt;em&gt;ω&lt;/em&gt;-DHFM model is well-suited for capturing the flow and heat transfer characteristics of liquid metal coolant in wire-wrapped fuel assemblies under buoyancy-dominated conditions. The predicted &lt;em&gt;Nu&lt;/em&gt; shows good agreement with both experimental data and empirical correlations, with deviations within acceptable limits. Furthermore, the characteristics of mixed and natural convection of liquid metals in a 19-pin wire-wrapped fuel assembly are investigated. In both mixed and natural convection of the coolant, the helical structure of the wire-wrapped significantly influences the temperature and velocity distributions. It not only induces vortex structures and periodic flow, but also creates local stagnation zones, leading to heat accumulation and the formation of hot spots. With the enhancement of buoyancy effects, flow variations primarily depend on temperature differences. Thermal plumes form a strong upward core flow, and the wire-wrapped geometry continues to exert a significant influence on the coolant flow. The SST &lt;em&gt;k&lt;/em&gt;-&lt;em&gt;ω&lt;/em&gt;-DHFM model is suitable for studying the flow and heat transfer characteristics of liquid metals under buoyancy effects, providing a numerical computation method for investigating the flow and heat transfer laws of liquid metals under buoyancy. Through this method, the mechanisms of flow and heat transfer of liquid metals under buoyancy effects have been revealed","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111761"},"PeriodicalIF":2.3,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750693","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 release of fission gas from defective cladding of PWR based on steady-state escape rate coefficient","authors":"Huaxiang Chen ,&nbsp;Bing Dong ,&nbsp;Junlian Yin ,&nbsp;Dezhong Wang ,&nbsp;Yuchen Song","doi":"10.1016/j.anucene.2025.111758","DOIUrl":"10.1016/j.anucene.2025.111758","url":null,"abstract":"<div><div>This study experimentally investigates the release behavior of fission gas through cladding defects in a pressurized water reactor, utilizing non-radioactive krypton as a surrogate for radioactive fission products. The results demonstrate a transient-to-steady evolution of gas release rates, characterized by an initial rapid surge followed by gradual decay to equilibrium. Minor defects tend to exhibit extended transient phases as a result of limited gas flux, whereas larger defects are characterized by elevated peak release rates along with abbreviated transient phases. Elevated temperature and pressure enhance peak release rates through intensified phase transitions. This process is modeled through a first-order kinetic equation incorporating the steady-state escape coefficient (<span><math><mrow><mi>ε</mi></mrow></math></span>). This coefficient is estimated via the Chapman-Enskog diffusion equation. Both experimental and theoretical results indicate that <span><math><mrow><mi>ε</mi></mrow></math></span> is in the order of 10⁻⁹ to 10⁻⁸ s⁻¹ for micrometer defects.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111758"},"PeriodicalIF":2.3,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738427","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":"Refined numerical analysis of prototype steam generator for sodium-cooled fast reactor","authors":"Bo Wang,&nbsp;Yan Zhang,&nbsp;Yang Yang,&nbsp;Zhengrong Guo,&nbsp;Jun Fan,&nbsp;Mingyang Li","doi":"10.1016/j.anucene.2025.111776","DOIUrl":"10.1016/j.anucene.2025.111776","url":null,"abstract":"<div><div>This paper presents a comprehensive three-dimensional numerical simulation of a prototype steam generator for a sodium-cooled fast reactor. The model features a 1:1 geometric representation of the sodium side with detailed meshing, while the water side utilizes a User-Defined Function for accurate thermal modeling. The simulation integrates the coupled flow and heat transfer processes on both sides of the sodium-water steam generator. Validation against experimental data confirms the model’s reliability, showing good agreement in sodium temperature and heat load distribution under both rated and 28% power conditions. The study reveals detailed flow and heat transfer characteristics, including sodium-side flow fields, pressure fields, and temperature fields, as well as water-side fluid and wall temperatures and heat transfer coefficients. Results indicate a relatively uniform heat load distribution in most areas of the tube bundle. This work provides a robust theoretical foundation and effective technical support for the design optimization, performance improvement, and safety assessment of sodium-cooled fast reactor steam generators.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111776"},"PeriodicalIF":2.3,"publicationDate":"2025-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144724721","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 temperature control technology of high-temperature irradiation testing of fuel","authors":"Wenhua Yang ,&nbsp;Yixiong Sun ,&nbsp;Liang Zhang,&nbsp;Shuai Jin,&nbsp;Wenlong Zhang,&nbsp;Sheng Sun","doi":"10.1016/j.anucene.2025.111724","DOIUrl":"10.1016/j.anucene.2025.111724","url":null,"abstract":"<div><div>Gas-cooled reactors, with their high operating temperatures, are key candidates for space nuclear power. Ensuring the reliability of new fuel elements necessitates high-temperature irradiation testing in research reactors. Multiple potential temperature control strategies, integrating existing technologies suitable for such testing, are proposed and analyzed. For the scheme based on the conductive heat transfer of prototype reactor coolant, the central temperature of fuel increases rapidly and then decreases slowly as the thickness of gas gap increases. When the thickness reaches a certain level, fuel temperature becomes insensitive to the changes in thickness. Conversely, a convective heat transfer-based scheme highlights gas velocity and inlet temperature as critical, with fuel temperature inversely related to velocity and directly to temperature, while keeping cladding temperatures lower than that of scheme based on conduction. An inert gas mixture scheme offers precise temperature control and emergency cooling during high-temperature tests, ensuring safety and adaptability for super-power operations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111724"},"PeriodicalIF":2.3,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144866376","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":"Oscillatory heat transfer analysis of assisting and opposing radiating flow of Williamson-nanofluid along heat-exchanger plate in nuclear-power reactors: numerical simulation","authors":"Zia Ullah ,&nbsp;Mohammed Alkinidri","doi":"10.1016/j.anucene.2025.111777","DOIUrl":"10.1016/j.anucene.2025.111777","url":null,"abstract":"&lt;div&gt;&lt;div&gt;Exothermic catalytic chemical reaction and thermal solar energy aspects for steady and fluctuating heat and mass flow rate using Williamson nanofluid is very important in nuclear power reactors. Main purpose of this work is to analyze viscous dissipation effects for fluid temperature field, concentration reactants, stead and oscillatory heat-mass transport, streamlines and isotherms along heat exchanger plate of nuclear power reactor using opposing and assisting flow. Significance of this study is to display oscillations and periodical heat transfer and skin friction using opposing and assisting flow. Mathematical formulation is developed for the computational assessment of laminar stream rate and fluctuating heat-mass transmission by using amplitude and phase change analysis. For programming algorithm and numerical results, the mathematical model is solved by dimensionless variables, Stokes oscillating factors, primitive transformations, implicit finite-difference methodology, and Gaussian elimination computation. For numerical results, the following range of parameters exothermic reaction rate &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;mi&gt;r&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;2.0&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Eckert number &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;E&lt;/mi&gt;&lt;mi&gt;c&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;2.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, radiation &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;mi&gt;d&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;6.0&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Prandtl 0.1 &lt;span&gt;&lt;math&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;/math&gt;&lt;/span&gt; Pr &lt;span&gt;&lt;math&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;/math&gt;&lt;/span&gt; 10.0, local-modified buoyancy forces &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;λ&lt;/mi&gt;&lt;mi&gt;T&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;λ&lt;/mi&gt;&lt;mi&gt;C&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;4.0&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Schmidt number &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;S&lt;/mi&gt;&lt;mi&gt;c&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;1.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, thermophoresis &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;N&lt;/mi&gt;&lt;mi&gt;T&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;0.8&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Brownian motion &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;N&lt;/mi&gt;&lt;mi&gt;B&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;0.6&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, Weissenberg number &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mi&gt;W&lt;/mi&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;2.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, and activation energy &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;0.1&lt;/mn&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;E&lt;/mi&gt;&lt;mi&gt;A&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;≤&lt;/mo&gt;&lt;mn&gt;1.5&lt;/mn&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; are considered. First steady flow rate is calculated and used in fluctuating formula to examine oscillatory flow rate. The percentage rate (28 %) of heat transfer is increased at small choice of Eckert number&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;E&lt;/mi&gt;&lt;mi&gt;c&lt;/mi&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; = 0.1. The maximum percentage (16 %) of mass rate is observed at high Eckert number&lt;span&gt;&lt;math&gt;&lt;msub&gt;&lt;mi&gt;E&lt;/mi&gt;&lt;mi&gt;c&lt;/mi&gt;&lt;/msub&gt;&lt;/math&gt;&lt;/span&gt; = 2.0. It is observed that streamlines contour increases as radiation-energy increases and exothermic/catalytic reaction flow decreases. It is deduc","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111777"},"PeriodicalIF":2.3,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144722642","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":"Small modular reactors in integrated energy systems: Load-following performance under power demand optimization","authors":"Jieling Wang ,&nbsp;Zelin You ,&nbsp;JiaWei Tan ,&nbsp;Yichun Wu ,&nbsp;Rui Jing","doi":"10.1016/j.anucene.2025.111748","DOIUrl":"10.1016/j.anucene.2025.111748","url":null,"abstract":"<div><div>The increasing consumption of fossil fuels and the growing severity of environmental pollution have become critical global issues. Small Modular Reactors (SMRs), recognized for their clean energy production and operational flexibility, have attracted significant interest for their potential integration into Integrated Energy Systems (IES). However, limited research has explored the load-following capabilities of SMRs in practical applications of Nuclear-Renewable Integrated Energy Systems (NR-IES) through simulation. To bridge this gap, this study develops a co-simulation framework that integrates a dynamic simulation model of SMRs with an optimization model of NR-IES to systematically assess the feasibility of SMRs in IES applications. Simulation results demonstrate that SMRs can effectively follow demand variations within 0%–50% of full power (FP). Their modular design and flexible operation significantly mitigate the intermittency of renewable energy sources, enhancing overall energy utilization efficiency and reducing dependence on energy storage. This study provides valuable insights for the design, operation, and evaluation of NR-IES, contributing to the advancement of sustainable and resilient energy systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111748"},"PeriodicalIF":2.3,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144866385","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 an intermediate break loss of coolant accident (IBLOCA) under the OPR1000 operation condition","authors":"Yusun Park,&nbsp;Byoung-Uhn Bae,&nbsp;Jae Bong Lee,&nbsp;Seok Cho,&nbsp;Jun Yeong Jung,&nbsp;Kyoung-Ho Kang,&nbsp;Seung Wook Lee","doi":"10.1016/j.anucene.2025.111730","DOIUrl":"10.1016/j.anucene.2025.111730","url":null,"abstract":"<div><div>In the safety analysis of current Design-Based Accidents (DBAs), the Loss of Coolant Accidents (LOCA) are divided into Large Break Loss of Coolant Accidents (LBLOCA) and Small Break Loss of Coolant Accidents (SBLOCA) depending on the break size. In Korea, a Deterministic Safety Analysis (DSA) methodology is applied to classify LOCAs according to the break size. Currently, Korea’s regulatory system also divides LOCA into large and small break LOCA depending on the break size, and does not consider Intermediate Break LOCA (IBLOCA). There has also been almost no research on IBLOCA conducted in Korea. Therefore, in order to evaluate the safety of IBLOCA, a study must first be conducted to evaluate the safety of IBLOCA for operating nuclear power plant and to evaluate the prediction capability of safety analysis codes for IBLOCA transient through the Integral Effect Tests (IETs). In this study, an integral effect test (IET) database was established by utilizing the ATLAS facility to verify SPACE code for an intermediate break loss of coolant accident (IBLOCA) application. The test was performed by simulating an IBLOCA under the operational condition of OPR1000 (Optimized Pressurized Reactor 1000MWe) nuclear power plant. During the transient simulation, the major thermal hydraulic phenomena that can be observed when the IBLOCA occurs in OPR1000 was observed. With the break initiation, the collapsed water level in the reactor pressure vessel decreased rapidly and the core heaters were exposed to the steam in a short time. However, the coolant was supplied to the primary system through continuous operation of the safety injection systems, and the whole system was stably cooled down despite assuming a single failure of each safety injection system. Based on this test result, the system cooling capability with an operation of safety systems during an IBLOCA transient was evaluated. And this test data was utilized to verify and evaluate the SPACE code for application to an IBLOCA transient.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111730"},"PeriodicalIF":2.3,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144866384","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 deterministic continuous-energy neutron transport calculation method based on hybrid basis function expansion","authors":"Haopo Liu,&nbsp;Yunzhao Li,&nbsp;Xiaoyu Wen,&nbsp;Hongchun Wu,&nbsp;Liangzhi Cao","doi":"10.1016/j.anucene.2025.111695","DOIUrl":"10.1016/j.anucene.2025.111695","url":null,"abstract":"<div><div>The Boltzmann neutron transport equation is capable of describing the neutron transport process. The deterministic methods for solving the equation have demonstrated their superior computational efficiency to stochastic methods. However, the treatment of the energy variable remains critically reliant on the multi-group approximation. Its inherent limitations contain a dependence on problem-specific representative energy spectra and the necessity for complex and problem-dependent resonance calculations. Accordingly, this work presents a novel continuous-energy deterministic neutron transport methodology based on basis function expansion for different energy ranges and their coupling effects. Neutron energy spectra are represented using different orthogonal basis sets: polynomial bases for non-resonant energy ranges and wavelet bases for resonant energy ranges. It enables the accurate resolution of complex spectral behavior across the entire energy domain and obviates the need for representative spectra and resonance calculation. Preliminary validation studies performed on a series of uniform medium problems demonstrate its feasibility and accuracy.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111695"},"PeriodicalIF":2.3,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144865787","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}