Progress in Nuclear Energy最新文献

{"title":"Induction heating of composite billet of T91 and Zircaloy-4 for manufacturing of double clad tube: Simulation and experimental validation","authors":"Navinkumar Dubey ,&nbsp;Aarti Chaudhary ,&nbsp;Arijit Laik ,&nbsp;Sugilal Gopalakrishnan ,&nbsp;Komal Kapoor","doi":"10.1016/j.pnucene.2025.105761","DOIUrl":"10.1016/j.pnucene.2025.105761","url":null,"abstract":"<div><div>Metallic fuel is the preferred choice for future Fast Breeder Reactors due to higher breeding ratio. A mechanical bonded double clad tube with a T91 steel outer shell and Zr inner liner is a potential choice for metallic fuel. However, the metallurgical bonded clad tube will have superior properties against the mechanical bonded tube. The metallurgically bonded co-axial tube of T91 and Zr/Zircaloy-4 (Zr-4) can be manufactured by the co-extrusion route. The significant difference in deformation behavior of the two alloys demands a temperature gradient over the thickness of the composite billet for co-extrusion. The fine control of temperature gradient across the cross section of the billet can be achieved only by induction heating. The present study focuses on simulating the induction preheating process of a double-clad billet of outer T91 and inner Zr-4. A 2-D axisymmetric model was considered and simulation studies with several key parameters such as coil frequency, current, pitch, and diameter were carried out with an aim to reduce temperature variation at the interface of T91 and Zr-4 in the billet. Simulations using optimised coil frequency and current reduced the temperature variation at the interface to as low as 22 °C. Subsequent simulations with variable coil pitch and diameter reduced the temperature variation to less than ±3 °C at the interface ignoring axial ends. The mean temperature across the T91 was 967.4 °C and across Zr-4 was 850.3 °C. Experiments were conducted for the co-axial tube of T91 and Zr-4 to validate the simulation model.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105761"},"PeriodicalIF":3.3,"publicationDate":"2025-03-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143725986","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":"Enhancing nuclear power plant diagnostics: A comparative analysis of XAI-based feature selection methods for abnormal and emergency scenario detection","authors":"Young Ho Chae ,&nbsp;Seung Geun Kim ,&nbsp;Jeonghun Choi ,&nbsp;Seo Ryong Koo ,&nbsp;Jonghyun Kim","doi":"10.1016/j.pnucene.2025.105759","DOIUrl":"10.1016/j.pnucene.2025.105759","url":null,"abstract":"<div><div>This study introduces the application of explainable artificial intelligence (XAI) techniques to enhance nuclear power plant diagnostics through effective feature selection. We compared various XAI methods, including gradient-based techniques, layer-wise relevance propagation, DeepSHAP, integrated gradients, local interpretable model-agnostic explanation(LIME), and saliency maps, with traditional approaches such as principal component analysis (PCA). By applying these methods to data from an IAEA iPWR simulator, which includes 35 abnormal and emergency scenarios with 116 state variables, we demonstrated the superiority of XAI-based methods in selecting features that effectively distinguish between different plant conditions. Our approach successfully reduced the input dimensionality from 116 to 20 features while maintaining high diagnostic accuracy. XAI methods, particularly saliency map and DeepSHAP, outperformed traditional techniques by revealing distinct patterns for various abnormal situations. This reduction in dimensionality offers several benefits, including enhanced cybersecurity, improved human–machine interfaces, and increased computational efficiency. The findings have significant implications for developing more accurate, efficient, and interpretable diagnostic systems in nuclear power plants, potentially improving safety and operational effectiveness. Future work will focus on validating these methods across diverse plant designs and integrating this approach with advanced AI techniques for real-time adaptive diagnostics.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105759"},"PeriodicalIF":3.3,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143704338","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Influence of healing time, technological void and temperature on the shear behavior of the GMZ bentonite healing interface","authors":"Yi-ran Zhang ,&nbsp;Yong-gui Chen ,&nbsp;Yu-cheng Li ,&nbsp;Wei-min Ye ,&nbsp;Qiong Wang ,&nbsp;Zhao Sun","doi":"10.1016/j.pnucene.2025.105757","DOIUrl":"10.1016/j.pnucene.2025.105757","url":null,"abstract":"<div><div>Bentonite is an engineering barrier in the deep geological repository (DGR), and its mechanical properties are negatively impacted by the presence of technological voids. Based on direct shear test and mercury intrusion porosimetry (MIP) technique, the effects of healing time, technological void and temperature on the healing interface shear behavior of Gaomiaozi (GMZ) bentonite were studied. The results indicate that the shear strength of healing interface is directly related to the dry density of the interface, which is influenced by healing time, technological void and temperature. Within the healing time range studied, as healing time increased, bentonite continued to swell, leading to a rise in the dry density of the interface and an enhancement in shear strength. During this process, constrained by the limited swelling space, the intra-aggregate porosity decreases at first and then increases until stabilizing, while the inter-aggregate porosity is on the contrary, which reduces the distance between particles, thereby increasing the cohesion and friction angle of the interface. When the initial void of the interface increases, the shear strength at the interface significantly decreases, with corresponding reductions in cohesion and friction angle. As the temperature increases, the shear strength increases, notably in cohesion, while the friction angle shows no clear pattern of change. This is primarily attributed to the rearrangement of bentonite particles, the reformation of interlayer physical and chemical bonds, and the reduced viscosity of pore fluids. The findings of this study provide important insights into the mechanical stability of engineered barriers in the DGR.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105757"},"PeriodicalIF":3.3,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143715068","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":"Evaluating the performance of passive second-stage hydro accumulators to enhance safety margins at Bushehr NPP","authors":"Mehdi Yarizadeh-Bene ,&nbsp;Mahdi Zangian ,&nbsp;Abdolhamid Minuchehr ,&nbsp;Hamed Kargaran","doi":"10.1016/j.pnucene.2025.105744","DOIUrl":"10.1016/j.pnucene.2025.105744","url":null,"abstract":"<div><div>Following the Fukushima accident, the nuclear industry acknowledged the significance of passive safety systems that may function autonomously without relying on external power sources or operator involvement. These systems are engineered to deliver reliable and automated safety operations, even in severe circumstances. The Second Stage Hydro accumulator (HA-2) is an autonomous safety system that autonomously introduces coolant into a nuclear reactor during emergencies, without the need for external power. It plays a vital role in the safety architecture of sophisticated VVER reactors. The objective of this study is to examine the influence of the new type of second-stage hydro accumulators (HA-2) compared to the second-stage accumulators (Bushehr) that are only in Bushehr NPP on the occurrence of design extension condition (DEC) during a station blackout and a small break in the coolant pipeline along with an SBO accident. For this purpose, a VVER-1000 thermal hydraulic model for RELAP5 code has been modified to include the HA-2 system. After analyzing the results, it has been concluded that the passive part of the safety system, together with the accumulators (ACCs), performs well enough to prevent core damage in small Loss of Coolant Accident (SB-LOCA) scenarios larger than 70 mm, as well as in Station Blackout (SBO) situations, for a minimum of 13 h. Furthermore, in circumstances including a severe blackout (SBO), the system exhibits adequate performance to prevent any damage to the core for 24 h. These results are based on assessing the system's autonomous capacity, without considering supplementary managerial interventions<strong>.</strong></div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105744"},"PeriodicalIF":3.3,"publicationDate":"2025-03-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143715067","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Research on leakage and migration behavior of 87Kr/135mXe in the Molten Salt Reactor containment","authors":"Honghua Peng ,&nbsp;Guifeng Zhu ,&nbsp;Zhihong Zhang ,&nbsp;Jun Cai ,&nbsp;Defeng Chen ,&nbsp;Qiang Qin ,&nbsp;Jinsen Xie","doi":"10.1016/j.pnucene.2025.105758","DOIUrl":"10.1016/j.pnucene.2025.105758","url":null,"abstract":"<div><div>Effective monitoring and control of radioactive gas leakage are crucial for the safe operation of Molten Salt Reactors (MSRs). The liquid fuel in MSRs increases the mobility of radioactive materials. Therefore, predicting the migration and leakage of volatile radioactive products from primary system boundaries is essential. This study, for the first time, focuses on the migration behavior of radioactive gases within the MSR containment during the first 2 h following the appearance of small leaks. Using the FLUENT software, the research investigates the influence of gas flow fields, leakage locations, and decay characteristics on the transport of these radioactive gases, particularly ⁸⁷Kr and ^135mXe. The results indicate that pre-existing gas flow fields significantly affect the distribution and migration of radioactive gases, leading to a more uniform concentration distribution. Furthermore, the concentration ratios of ⁸⁷Kr to ^135mXe exhibit a near-linear increase with leakage time, with the specific magnitude of increase being strongly correlated with leakage locations. When the initial leaking concentration ratio of ⁸⁷Kr to ^135mXe is set at approximately 1.70, the concentration ratios for leakage locations (i.e., point 1, point 2, and point 3) increase to 1.81, 1.86, and 1.89 at 3 min after leakage, and to 5.90, 6.80, and 7.94 at 2 h after leakage, respectively. This study proposes a novel monitoring approach based on the concentration and concentration ratios of ⁸⁷Kr and ^135mXe for MSRs. This approach provides valuable information for leakage diagnosis, such as leakage time, location, and flow rate, and offers technical support for the radiation monitoring and safe operation of MSRs.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105758"},"PeriodicalIF":3.3,"publicationDate":"2025-03-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143687721","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 analysis of a large two phase fuel bubble evolution under energetic CDA in a medium-sized pool type SFR","authors":"B. Thilak ,&nbsp;P. Mangarjuna Rao ,&nbsp;Jose Varghese ,&nbsp;B.K. Sreedhar","doi":"10.1016/j.pnucene.2025.105729","DOIUrl":"10.1016/j.pnucene.2025.105729","url":null,"abstract":"<div><div>The two-phase fuel bubble behavior inside the liquid sodium pool under energetic core disruptive accident (CDA) is an essential consideration in the safety analysis of sodium fast reactors (SFR). The fuel bubble expansion and the resultant reactor vessel pressurization during energetic CDA drive the enhanced actinide transport from the damaged core to the cover gas region and the liquid sodium release to the reactor containment building (RCB). A numerical model is developed to evaluate the fuel bubble behavior by considering the sodium entrainment and the fuel-sodium heat transfer in a medium-sized pool type SFR. The model is validated with relevant benchmark experimental results. Under energetic CDA conditions, the fuel bubble mass typically varies between 1000 and 3000 kg, and the initial fuel bubble temperature ranges from 4200 to 4700 K. The parametric analysis with these input values shows that the sodium entrainment at the bubble-pool interface and the fuel bubble behavior are more sensitive to the initial fuel bubble temperature or superheat than the fuel bubble mass. The time period for the fuel bubble's first expansion-compression cycle range between 600 and 800 ms. The equilibrium or quasi-static pressure under adiabatic condition (i.e. without sodium entrainment and fuel-sodium heat transfer) for the 1000 kg-4200 K case is 1.57 × 10⁵ Pa. Results also show that the sodium vapor partial pressure dominates the total fuel bubble pressure in the 1000 kg-4200 K case when the sodium entrainment and the fuel-sodium heat transfer effects are included. As a result, the quasi-static pressure in the reactor vessel increases to 3.0 × 10⁵ Pa. The fuel bubble-sodium pool interface velocity history which dictates the actinide upward displacement from the core region is evaluated as a function of initial fuel bubble temperature and mass. The maximum fuel bubble-sodium pool interface velocity for the input values considered is 35 ms⁻¹. A mechanistic evaluation of the reactor vessel pressure history after the completion of bubble expansion shows that the radiation heat transfer and the fuel vapor condensation reduce the reactor vessel pressure from the quasi-static to the ambient pressure within 10 s. The evaluated velocity and pressure histories would serve as inputs for the mechanistic estimation of in-vessel and in-containment accident source terms due to the fuel bubble expansion under energetic CDA scenario in SFR.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105729"},"PeriodicalIF":3.3,"publicationDate":"2025-03-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143687127","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 Euratom NPHyCo project: Conceptualization, technical work plan and current status","authors":"Eduard Diaz-Pescador ,&nbsp;Marco Viebach ,&nbsp;Wolfgang Lippmann ,&nbsp;Florian Gamaleja ,&nbsp;Antonio Hurtado ,&nbsp;Oleksandr Mazurok ,&nbsp;Cecilia Herrero-Moriana ,&nbsp;Martin Glückler ,&nbsp;Stéphanie Crevon ,&nbsp;Pavla Lukášová ,&nbsp;Michael A. Fütterer ,&nbsp;Mathijs Hijlkema ,&nbsp;Gabriele Firpo ,&nbsp;Jessica Johnson ,&nbsp;Clemens Heitsch","doi":"10.1016/j.pnucene.2025.105731","DOIUrl":"10.1016/j.pnucene.2025.105731","url":null,"abstract":"<div><div>This manuscript provides an overview of the goals, technical content, and close-to-final status of the Euratom NPHyCo project. NPHyCo stands for <u>N</u>uclear <u>P</u>owered <u>Hy</u>drogen <u>Co</u>generation, and its ultimate goal is the design and preparation of a large-scale (MW-range) hydrogen cogeneration facility coupled to a nuclear power plant (NPP), which can become operational in a short-term horizon. The project analyses the existing economic, technical, safety and regulatory considerations for NPP owners to produce nuclear hydrogen through several cases of study applied to selected NPP sites and European markets.</div><div>The manuscript summarizes the main outcomes as of December 2024 regarding the integration scenarios of a hydrogen production plant (HPP) into a NPP site, based on the coupling configuration and sharing of balance-of-plant components between facilities. Given the interest and degree of cooperation of the Ukrainian state operator Energoatom, most of the technical calculations are based on the specifications of Rivne and Khmelnytskyi NPPs, both powered with water-water energy reactor (VVER) technology.</div><div>The project has delineated a HPP configuration with an overall capacity of 45 MW. The safety assessment relies on deterministic and probabilistic methodologies, which are applied through numerical simulation tools. The results are mostly based on example scenarios. However, they allow to derive general conclusions. Deterministic calculations are applied to a hydrogen plant either in a container or building solution, to determine the optimal configuration and required safety measures upon an intended release of hydrogen and subsequent ignition, leading to a fire or explosion. The hazards of the worst-case scenario in the HPP towards safety-related NPP structures are evaluated based on onsite and offsite HPP integration at Rivne NPP, with the corresponding structural fragility criterion.</div><div>The techno-economic study addresses the hydrogen value chain, encompassing low-temperature and high-temperature electrolysis technologies, up to storage and transportation. The study evaluates if existing systems and resources at the NPP are sufficient and suitable to supply the HPP, and thereby lead to reduced capital expenses (CAPEX) and operational expenses (OPEX). A summary of techno-economic analysis is summarized in this paper. Several sensitivity analyses have served to obtain the configuration to minimize the levelized cost of hydrogen (LCOH) and advance operational strategies to determine the best dynamic configuration for the hybrid system operation. A final business plan for hydrogen production by means of already existing NPPs, has been also developed. For the licensing of the HPP in the vicinity of the NPP, the degree of integration and distance between facilities are the most critical factors. The licensing impact is investigated for three different levels of plant integration.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105731"},"PeriodicalIF":3.3,"publicationDate":"2025-03-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143687128","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experiments on countercurrent flow in a vertical pipe through perforated plate","authors":"Jie Wan ,&nbsp;Xin Ma ,&nbsp;Huafa Chen ,&nbsp;Wan Sun ,&nbsp;Yang Liu ,&nbsp;Ren Liang ,&nbsp;Longxiang Zhu ,&nbsp;Luteng Zhang ,&nbsp;Qiang Lian ,&nbsp;Simiao Tang ,&nbsp;Liang-ming Pan","doi":"10.1016/j.pnucene.2025.105730","DOIUrl":"10.1016/j.pnucene.2025.105730","url":null,"abstract":"<div><div>Gas-liquid countercurrent flow through perforated plate is of great significance for the safety and efficient operation of industrial systems. In this study, visual observations of flow patterns during countercurrent flow through a perforated plate in a vertical pipe was presented. The water penetration flow rate measured by the weighing method, along with the flow patterns, was employed as a criterion for the onset of flooding. The effects of the height above the perforated plate and the structure of the perforated plate on flooding were studied. The results revealed two mechanisms for flooding under different height above the perforated plate, which can be characterized by the Froude number (Fr). For perforated plates with a larger number of holes, the wetted perimeter increases, resulting in a more uniform liquid distribution across the plate. This further leads to a higher gas flow rate required for the flooding. Additionally, the effect of the shape of hole including circular and square on the flooding was investigated. Furthermore, the fraction of liquid carryover was determined by the measured liquid penetration flow rate and injection flow rate. Considering the effects of gas flow rate, the height above the perforated plate and the structure of the perforated plate which is described by the wetted perimeter, a model for predicting the fraction of liquid carryover in countercurrent flow through perforated plates was established.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105730"},"PeriodicalIF":3.3,"publicationDate":"2025-03-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143687129","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 use of ferro frit 3195 to stably solidify oxide sludge generated by corrosion of metallic structure in nuclear power plant","authors":"Ki Joon Kang,&nbsp;Sia Hwang,&nbsp;Hee Reyoung Kim","doi":"10.1016/j.pnucene.2025.105746","DOIUrl":"10.1016/j.pnucene.2025.105746","url":null,"abstract":"<div><div>Sludge waste was mixed with ferro frit powder, a sintering agent, and the homogeneous and inhomogeneous distributions of the sintering agent in the sludge were compared and analyzed. In nuclear power plants containing stainless-steel disposal tanks, stainless-steel oxide-based sludge is generated from long-term exposure to radioactive fluid waste, and this sludge waste should be disposed of via solidification. By adding only 10 % ferro frit 3195 to 90 % oxide sludge waste, stable solidified sludge was made. The production processes of solidified sludge consisted of three parts: homogeneous mixing of ferro frit 3195 and sludge, compression molding of the mixed powder by hydraulic press at room temperature, and heating at 900 <span><math><mrow><mo>°C</mo></mrow></math></span>, where the solidified body with the homogeneous distribution of ferro frit 3195 could form stable solidified sludge. The leachability index of nonradioactive Co and the compressive strength of stable solidified sludge made by homogeneous mixing of ferro frit and oxide sludge were 16.9 <span><math><mrow><mo>±</mo></mrow></math></span> 0.0647 and 4.97 MPa, respectively.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105746"},"PeriodicalIF":3.3,"publicationDate":"2025-03-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143687130","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Research of the depletion calculation method based on the core nuclear design software package TORCH","authors":"Chen Zhao,&nbsp;Bin Zhang,&nbsp;Shien Huang,&nbsp;Zhang Chen,&nbsp;Xiaoming Chai","doi":"10.1016/j.pnucene.2025.105747","DOIUrl":"10.1016/j.pnucene.2025.105747","url":null,"abstract":"<div><div>In the two-step method of reactor physics design calculation, depletion calculation is one of the most important modules for accuracy and efficiency. The depletion method has been researched in two aspects, including the micro-depletion and the macro-depletion. These researches were implemented in the TORCH software package and validated with actual measured data from Qinshan, Daya bay and Fuqing nuclear power plants. Numerical results of critical boron concentration and power peaking factor show that the calculation accuracy of predictor-corrector method, improved predictor-corrector method and logarithmic linear interpolation method is gradually improved in micro-depletion method and the critical boron concentration is underestimated by 20 ppm in macro-depletion method. All results show reasonable deviations and new options for the depletion calculation can be provided in these methods.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"185 ","pages":"Article 105747"},"PeriodicalIF":3.3,"publicationDate":"2025-03-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143687720","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}