Annals of Nuclear Energy最新文献

英文中文

Magnetohydrodynamic conjugate mixed convection, Joule Heating, and entropy generation through a ferrofluid filled T-shaped open miniature chamber with a Heat-Generating circular rod

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-03-03 DOI: 10.1016/j.anucene.2025.111294

Md Tanbirul Islam Rupam , Nahid Hasan , Md. Sheikh Rasel , Sumon Saha

The present study computationally investigates magnetohydrodynamic (MHD) mixed convective fluid circulation and entropy generation in a T-shaped open chamber containing a heat-generating and conducting cylinder. Ferrofluid is circulated through the enclosure by entering at the bottom and leaving from the top of both side openings. This study utilizes the finite element scheme to unravel the leading thermal energy and Navier-Stokes equations, employing suitable auxiliary conditions. This research aims to analyze the effects of governing non-dimensional governing and geometric parameters and explore the best thermo-fluid performance inside the enclosure. The geometrical and controlling parameters are the cylinder location in the vertical direction (δ = 0.6, 0.7, 0.8), Reynolds number (31.62 ≤ Re ≤ 316.23), Grashof number (10³ ≤ Gr ≤ 10⁵), Richardson number (0.1 ≤ Ri ≤ 10), Stuart number (0 ≤ N ≤ 3.16), Hartmann number (0 ≤ Ha ≤ 17.78), and Joule heating parameter (0 ≤ J ≤ 4.57 × 10⁻⁸). The outcomes of this investigation are assessed using numerical computations of the overall entropy generation within the enclosure, average Nusselt number along the edge of the heated cylinder, mean temperature of the solid cylinder, and thermal performance criterion for six distinct cases. Furthermore, a visual depiction of the fluid circulation and thermal fields is presented. Upon thorough examination, it becomes evident that elevated Reynolds and Grashof numbers result in increased heat transport and reduced entropy production. Moreover, the optimal vertical location of the cylinder is identified at 0.6 times the chamber height. The maximum Nusselt number is achieved in Case 1 (at fixed N and Gr), where a 26.78 % improvement can be obtained by adjusting the parameter values at δ = 0.6. The inclusive discoveries of the current study grasp the noteworthy potential for apprising the design of miscellaneous thermal systems, together with solar thermal collectors, nuclear reactor cooling, electronic cooling, etc.

{"title":"Magnetohydrodynamic conjugate mixed convection, Joule Heating, and entropy generation through a ferrofluid filled T-shaped open miniature chamber with a Heat-Generating circular rod","authors":"Md Tanbirul Islam Rupam , Nahid Hasan , Md. Sheikh Rasel , Sumon Saha","doi":"10.1016/j.anucene.2025.111294","DOIUrl":"10.1016/j.anucene.2025.111294","url":null,"abstract":"<div><div>The present study computationally investigates magnetohydrodynamic (MHD) mixed convective fluid circulation and entropy generation in a <em>T</em>-shaped open chamber containing a heat-generating and conducting cylinder. Ferrofluid is circulated through the enclosure by entering at the bottom and leaving from the top of both side openings. This study utilizes the finite element scheme to unravel the leading thermal energy and Navier-Stokes equations, employing suitable auxiliary conditions. This research aims to analyze the effects of governing non-dimensional governing and geometric parameters and explore the best thermo-fluid performance inside the enclosure. The geometrical and controlling parameters are the cylinder location in the vertical direction (<em>δ</em> = 0.6, 0.7, 0.8), Reynolds number (31.62 ≤ <em>Re</em> ≤ 316.23), Grashof number (10<sup>3</sup> ≤ <em>Gr</em> ≤ 10<sup>5</sup>), Richardson number (0.1 ≤ <em>Ri</em> ≤ 10), Stuart number (0 ≤ <em>N</em> ≤ 3.16), Hartmann number (0 ≤ <em>Ha</em> ≤ 17.78), and Joule heating parameter (0 ≤ <em>J</em> ≤ 4.57 × 10<sup>−8</sup>). The outcomes of this investigation are assessed using numerical computations of the overall entropy generation within the enclosure, average Nusselt number along the edge of the heated cylinder, mean temperature of the solid cylinder, and thermal performance criterion for six distinct cases. Furthermore, a visual depiction of the fluid circulation and thermal fields is presented. Upon thorough examination, it becomes evident that elevated Reynolds and Grashof numbers result in increased heat transport and reduced entropy production. Moreover, the optimal vertical location of the cylinder is identified at 0.6 times the chamber height. The maximum Nusselt number is achieved in Case 1 (at fixed <em>N</em> and <em>Gr</em>), where a 26.78 % improvement can be obtained by adjusting the parameter values at <em>δ</em> = 0.6. The inclusive discoveries of the current study grasp the noteworthy potential for apprising the design of miscellaneous thermal systems, together with solar thermal collectors, nuclear reactor cooling, electronic cooling, etc.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111294"},"PeriodicalIF":1.9,"publicationDate":"2025-03-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143529812","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}

引用次数: 0

Assessment of nuclear desalination in a small modular reactor using membrane distillation

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-03-01 DOI: 10.1016/j.anucene.2025.111279

Gabriel C.G.R. da Silva , Carolina P. Naveira-Cotta , Kleber M. Lisboa , Renato M. Cotta , Jian Su

Despite the proven feasibility and cost-effectiveness of nuclear desalination in small modular reactors (SMRs), the exclusive use of waste heat for this purpose remains virtually unexplored. This work investigates coupling an SMR (NuScale) to a direct contact membrane distillation (DCMD) desalination plant with heat recovery and feed recycle. Both the reactor waste heat and the low-pressure (LP) steam extraction from the reactor turbine were considered as heat sources. The DCMD hollow fiber module was modeled as a porous medium with satisfactory accuracy. Key parameters affecting system performance, identified through factorial analysis, include membrane porosity, module length, feed superficial velocity, and fiber inner radius. Single and multiobjective optimization analyses revealed the feasibility of producing up to 3,810 m³/d of water without any reactor power loss, and up to 8,832 m³/d, with a 2.28 MWe power loss, using steam extraction. DCMD demonstrated competitiveness, especially when the reactor’s primary purpose is electricity generation.

{"title":"Assessment of nuclear desalination in a small modular reactor using membrane distillation","authors":"Gabriel C.G.R. da Silva , Carolina P. Naveira-Cotta , Kleber M. Lisboa , Renato M. Cotta , Jian Su","doi":"10.1016/j.anucene.2025.111279","DOIUrl":"10.1016/j.anucene.2025.111279","url":null,"abstract":"<div><div>Despite the proven feasibility and cost-effectiveness of nuclear desalination in small modular reactors (SMRs), the exclusive use of waste heat for this purpose remains virtually unexplored. This work investigates coupling an SMR (NuScale) to a direct contact membrane distillation (DCMD) desalination plant with heat recovery and feed recycle. Both the reactor waste heat and the low-pressure (LP) steam extraction from the reactor turbine were considered as heat sources. The DCMD hollow fiber module was modeled as a porous medium with satisfactory accuracy. Key parameters affecting system performance, identified through factorial analysis, include membrane porosity, module length, feed superficial velocity, and fiber inner radius. Single and multiobjective optimization analyses revealed the feasibility of producing up to 3,810 m<sup>3</sup>/d of water without any reactor power loss, and up to 8,832 m<sup>3</sup>/d, with a 2.28 MWe power loss, using steam extraction. DCMD demonstrated competitiveness, especially when the reactor’s primary purpose is electricity generation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111279"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520184","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}

引用次数: 0

A coordinate transformation method to simulate non-uniform radial deformation of nuclear reactor cores

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-02-28 DOI: 10.1016/j.anucene.2025.111292

Evgeny Nikitin, Emil Fridman

This paper presents the development of a neutronics modeling methodology for handling non-uniform mesh deformations in a deterministic 3D reactor core dynamics simulator. Coordinate transformation is applied to solve physical problems in deformed geometries, while using a numerical solver originally derived for regular geometries. Specifically, this paper introduces and evaluates a method based on coordinate transformation to simulate non-uniform deformations of nuclear reactor cores and their impact on the reactor physics behavior.

The method, developed for the 3D reactor core simulator DYN3D, is applicable to any nodal diffusion-based reactor physics solver. It was tested in four stages using sodium-cooled fast reactors: (1) proof of concept on a simplified core; (2) feasibility of modelling flowering scenarios in a realistic core; (3) performance evaluation through the Phénix reactor core flowering experiment; and (4) applicability to dynamic deformation scenarios, such as pressure wave propagation.

引用次数: 0

Evaluation of scale-up capability of best estimate code application on China advanced Gen-III reactor

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-02-28 DOI: 10.1016/j.anucene.2025.111305

Ye Yang , Qian Ye , Mengyan Hu , Xueyan Zhang , Jun Yang

The Code Scaling, applicability, and Uncertainty method states the code’s capability to scale up processes from test facilities to full-scale nuclear power plants needs to be validated and evaluated. The reason for this validation is that it is infeasible (or cost prohibitive) to perform meaningful experiments at full scale and the ability of numerical tools designed to simulate the performance of nuclear reactors can be proven only at reduced scale.

ACME is an integral test facility, which is designed to study the behavior of China Advanced Pressurized Water Reactor (PWR) under accident conditions. The RELAP5 code is the best estimate thermal hydraulic system code for performing nuclear power plant safety analysis. This study validates the code scale up capability for application on China Advanced PWR. Firstly, we propose a new evaluation scheme, which is to take the realistically constructed test facility as a reference and scale up its numerical model to the size of a prototype power plant strictly according to scaling laws. This method, on one hand, ensures that the numerical model of the test facility and the scale up numerical model maintain consistency in node division. On the other hand, it avoids the influence of engineering deviations. Secondly, a numerical model for the prototype power plant scale was established based on the ideal scaling laws. After that, a 2-inch cold leg break accident test was simulated on two different scale numerical models, and the calculation results were compared with experimental data. The RELAP5 scale up capability to predict the accident phenomenon of China Advanced Gen-III PWR was evaluated using both qualitative and Fast Fourier Transform Based Method (FFTBM) quantitative methods.

{"title":"Evaluation of scale-up capability of best estimate code application on China advanced Gen-III reactor","authors":"Ye Yang , Qian Ye , Mengyan Hu , Xueyan Zhang , Jun Yang","doi":"10.1016/j.anucene.2025.111305","DOIUrl":"10.1016/j.anucene.2025.111305","url":null,"abstract":"<div><div>The Code Scaling, applicability, and Uncertainty method states the code’s capability to scale up processes from test facilities to full-scale nuclear power plants needs to be validated and evaluated. The reason for this validation is that it is infeasible (or cost prohibitive) to perform meaningful experiments at full scale and the ability of numerical tools designed to simulate the performance of nuclear reactors can be proven only at reduced scale.</div><div>ACME is an integral test facility, which is designed to study the behavior of China Advanced Pressurized Water Reactor (PWR) under accident conditions. The RELAP5 code is the best estimate thermal hydraulic system code for performing nuclear power plant safety analysis. This study validates the code scale up capability for application on China Advanced PWR. Firstly, we propose a new evaluation scheme, which is to take the realistically constructed test facility as a reference and scale up its numerical model to the size of a prototype power plant strictly according to scaling laws. This method, on one hand, ensures that the numerical model of the test facility and the scale up numerical model maintain consistency in node division. On the other hand, it avoids the influence of engineering deviations. Secondly, a numerical model for the prototype power plant scale was established based on the ideal scaling laws. After that, a 2-inch cold leg break accident test was simulated on two different scale numerical models, and the calculation results were compared with experimental data. The RELAP5 scale up capability to predict the accident phenomenon of China Advanced Gen-III PWR was evaluated using both qualitative and Fast Fourier Transform Based Method (FFTBM) quantitative methods.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111305"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520182","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}

引用次数: 0

PSA level 1 analysis by considering redundancy and diversity of emergency diesel generators for APR-1400

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-02-28 DOI: 10.1016/j.anucene.2025.111302

Muhammad Zubair, Ronak Shahin Radkiany, Mahra Mohammad Aljaradi, Shamma Abdulla Alblooshi, Horeya Sultan Alhammadi, Yumna Akram

This paper builds upon the initial Level 1 Probabilistic Safety Assessment (PSA) for Emergency Diesel Generator (EDG) failures in the APR-1400 Nuclear Power Plant (NPP). By incorporating more complex EDG failure scenarios and expanding the scope of fault trees and event trees, it provides a more comprehensive safety performance analysis using AIMS-PSA software. The results highlight the reactor’s safety systems’ resilience in the face of an increasing number of malfunctioning EDGs. An in-depth understanding of the interdependencies and failure probability of the Essential Chilled Water System (ECWS), Component Cooling Water System (CCWS), 480 V bus, and fault trees for the 125 V and 4.16 kV buses has been obtained through this detailed examination. The Core Damage Frequency (CDF) values from the analysis, 1.71E-6 /RY, were found to be significantly lower compared to official literature values of 2.56E-6 /RY and 2.25E-6 /RY, attributed to the exclusion of more extensive operational and environmental aspects of LOOP and common cause failure events. The findings provide critical insights for enhancing nuclear safety frameworks and underscore the need for scenario-based PSAs to address potential vulnerabilities. Overall, the paper not only effectively measures the resilience of the NPP’s safety measures to various failure modes but also provides regulators and plant management with insightful information about the significance of redundancy and diversity in preserving nuclear safety over the NPP’s operational lifetime.

{"title":"PSA level 1 analysis by considering redundancy and diversity of emergency diesel generators for APR-1400","authors":"Muhammad Zubair, Ronak Shahin Radkiany, Mahra Mohammad Aljaradi, Shamma Abdulla Alblooshi, Horeya Sultan Alhammadi, Yumna Akram","doi":"10.1016/j.anucene.2025.111302","DOIUrl":"10.1016/j.anucene.2025.111302","url":null,"abstract":"<div><div>This paper builds upon the initial Level 1 Probabilistic Safety Assessment (PSA) for Emergency Diesel Generator (EDG) failures in the APR-1400 Nuclear Power Plant (NPP). By incorporating more complex EDG failure scenarios and expanding the scope of fault trees and event trees, it provides a more comprehensive safety performance analysis using AIMS-PSA software. The results highlight the reactor’s safety systems’ resilience in the face of an increasing number of malfunctioning EDGs. An in-depth understanding of the interdependencies and failure probability of the Essential Chilled Water System (ECWS), Component Cooling Water System (CCWS), 480 V bus, and fault trees for the 125 V and 4.16 kV buses has been obtained through this detailed examination. The Core Damage Frequency (CDF) values from the analysis, 1.71E-6 /RY, were found to be significantly lower compared to official literature values of 2.56E-6 /RY and 2.25E-6 /RY, attributed to the exclusion of more extensive operational and environmental aspects of LOOP and common cause failure events. The findings provide critical insights for enhancing nuclear safety frameworks and underscore the need for scenario-based PSAs to address potential vulnerabilities. Overall, the paper not only effectively measures the resilience of the NPP’s safety measures to various failure modes but also provides regulators and plant management with insightful information about the significance of redundancy and diversity in preserving nuclear safety over the NPP’s operational lifetime.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111302"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520183","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}

引用次数: 0

Effect of CdS/M doping on the dielectric and radiation shielding of PVA/PEG composite film

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-02-28 DOI: 10.1016/j.anucene.2025.111321

Mohamed Bakr Mohamed , A.M. El-naggar , Zein K. Heiba , A.M. Kamal

The present study seeks to synthesize CdS and CdS/M (M = Fe, Mg, Mn) nanoparticles incorporated into a polyvinyl alcohol (PVA)/PEG (polyethylene glycol) blend polymer film, with the objective of developing an advanced nanocomposite for diverse applications in energy storage and radiation shielding. The phases formed in the fillers samples were identified using the X-ray diffraction technique. The structural, morphology, dielectric, and radiation shielding features of the films containing the nanofiller were assessed. Doped blend with CdS/Fe demonstrated the highest dielectric constant and ac conductivity values. The doped blend including CdS has the greatest energy density values. The relaxation time affected by the kind of filler. The influence of type of filler in host blended polymer on linear attenuation coefficients (LAC), mean free path (MFP), mass attenuation coefficient (MAC), half value length (HVL), tenth value length (TVL), equivalent atomic number (Z_eq), effective atomic number (Z_e_ff), electron density (N_eff), atomic cross-section (ACS), and electronic cross-section (ECS), fast neutron removal cross-section (FNRCS), buildup factor values was examined using Phy-X/PSD program. The obtained results demonstrated that doped blends displayed a propensity to interact with gamma rays instead of merely transmitting through them. Our doped blends exhibit superior neutron shielding properties compared to the undoped PVA/PEG blended polymer. The results indicated that doped blends may serve as viable and interesting nanocomposites for radiation shielding application. Also, CdS/Fe sample can be used in the energy storage capacitance application.

{"title":"Effect of CdS/M doping on the dielectric and radiation shielding of PVA/PEG composite film","authors":"Mohamed Bakr Mohamed , A.M. El-naggar , Zein K. Heiba , A.M. Kamal","doi":"10.1016/j.anucene.2025.111321","DOIUrl":"10.1016/j.anucene.2025.111321","url":null,"abstract":"<div><div>The present study seeks to synthesize CdS and CdS/M (<em>M</em> = Fe, Mg, Mn) nanoparticles incorporated into a polyvinyl alcohol (PVA)/PEG (polyethylene glycol) blend polymer film, with the objective of developing an advanced nanocomposite for diverse applications in energy storage and radiation shielding. The phases formed in the fillers samples were identified using the X-ray diffraction technique. The structural, morphology, dielectric, and radiation shielding features of the films containing the nanofiller were assessed. Doped blend with CdS/Fe demonstrated the highest dielectric constant and ac conductivity values. The doped blend including CdS has the greatest energy density values. The relaxation time affected by the kind of filler. The influence of type of filler in host blended polymer on linear attenuation coefficients (LAC), mean free path (MFP), mass attenuation coefficient (MAC), half value length (HVL), tenth value length (TVL), equivalent atomic number (<em>Z</em><sub>eq</sub>), effective atomic number (<em>Z<sub>e</sub></em><sub>ff</sub>), electron density (<em>N<sub>eff</sub>)</em>, atomic cross-section (ACS), and electronic cross-section (ECS), fast neutron removal cross-section (FNRCS), buildup factor values was examined using Phy-X/PSD program. The obtained results demonstrated that doped blends displayed a propensity to interact with gamma rays instead of merely transmitting through them. Our doped blends exhibit superior neutron shielding properties compared to the undoped PVA/PEG blended polymer. The results indicated that doped blends may serve as viable and interesting nanocomposites for radiation shielding application. Also, CdS/Fe sample can be used in the energy storage capacitance application.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111321"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520181","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}

引用次数: 0

Development and application of an external cooling model for pressure vessels based on the ISAA program

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-02-28 DOI: 10.1016/j.anucene.2025.111315

Runze Zhai , Bin Zhang , Jingliang Zhang , Shaowei Tang , Jianqiang Shan

The integrity of the pressure vessel is crucial to the safety of a reactor. During severe accidents in nuclear power plants, the IVR-ERVC (In-Vessel Retention-External Reactor Vessel Cooling) strategy is commonly adopted to prevent lower head failure. The heat transfer coefficient of the external coolant directly determines the success of the IVR-ERVC strategy. This study examines the IVR-ERVC strategy using the Integrated Severe Accident Analysis Program (ISAA). The ISAA program originally considers only three heat transfer modes for external cooling of the reactor lower head, with an overly conservative criterion for transition boiling. To improve the assessment of lower head failure, we developed a more accurate external cooling model. The model includes all external heat transfer modes. Its accuracy was validated by comparing the results with the ULPU-LIKE experimental data, with the maximum error in lower head wall temperature not exceeding 5 %. Finally, we applied the improved ISAA program to analyze a severe accident triggered by a Station Blackout (SBO) in a nuclear power plant, investigating the effectiveness of the IVR strategy for second-generation pressurized water reactors. The results indicate that in the event of an SBO accident, penetration component failure occurs at the 0° position on the lower head bottom after 8.03 h.

{"title":"Development and application of an external cooling model for pressure vessels based on the ISAA program","authors":"Runze Zhai , Bin Zhang , Jingliang Zhang , Shaowei Tang , Jianqiang Shan","doi":"10.1016/j.anucene.2025.111315","DOIUrl":"10.1016/j.anucene.2025.111315","url":null,"abstract":"<div><div>The integrity of the pressure vessel is crucial to the safety of a reactor. During severe accidents in nuclear power plants, the IVR-ERVC (In-Vessel Retention-External Reactor Vessel Cooling) strategy is commonly adopted to prevent lower head failure. The heat transfer coefficient of the external coolant directly determines the success of the IVR-ERVC strategy. This study examines the IVR-ERVC strategy using the Integrated Severe Accident Analysis Program (ISAA). The ISAA program originally considers only three heat transfer modes for external cooling of the reactor lower head, with an overly conservative criterion for transition boiling. To improve the assessment of lower head failure, we developed a more accurate external cooling model. The model includes all external heat transfer modes. Its accuracy was validated by comparing the results with the ULPU-LIKE experimental data, with the maximum error in lower head wall temperature not exceeding 5 %. Finally, we applied the improved ISAA program to analyze a severe accident triggered by a Station Blackout (SBO) in a nuclear power plant, investigating the effectiveness of the IVR strategy for second-generation pressurized water reactors. The results indicate that in the event of an SBO accident, penetration component failure occurs at the 0° position on the lower head bottom after 8.03 h.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111315"},"PeriodicalIF":1.9,"publicationDate":"2025-02-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143512699","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}

引用次数: 0

Impact on the amount of high-level waste of minor actinide in combination with americium transmutation and TRU generation reduction UO2 fuel in BWR

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-02-27 DOI: 10.1016/j.anucene.2025.111308

Kouji Hiraiwa , Rei Kimura , Satoshi Wada , Tsukasa Sugita , Kenichi Yoshioka , Satoshi Takeda , Takanori Kitada

We performed americium transmutation in uranium fuel with a transuranic generation reduction fuel, which uses BWR fuel and increases the ²³⁵U enrichment within the range of High-Assay Low-Enriched Uranium (HALEU). We evaluated the relationship between the total decay heat of ²⁴¹Am and ²⁴⁴Cm. The increase in decay heat from ²⁴⁴Cm when the ²³⁵U enrichment is 3.8 wt%, which is the example case of LEU application as that of existing fuel. The generation of ²⁴⁴Cm comes from a smaller mass number of TRU nuclides decrease as the neutron flux and neutron capture reactions decrease with increasing enrichment. As a result, in the range where the enrichment exceeds about 7.5 wt%, and the ratio of the decrease in the weight fraction of ²⁴¹Am to the increase in the weight fraction of ²⁴⁴Cm decreases below 4.1 %, and the total decay heat of the two nuclides is lower than it would be in no-transmutation.

{"title":"Impact on the amount of high-level waste of minor actinide in combination with americium transmutation and TRU generation reduction UO2 fuel in BWR","authors":"Kouji Hiraiwa , Rei Kimura , Satoshi Wada , Tsukasa Sugita , Kenichi Yoshioka , Satoshi Takeda , Takanori Kitada","doi":"10.1016/j.anucene.2025.111308","DOIUrl":"10.1016/j.anucene.2025.111308","url":null,"abstract":"<div><div>We performed americium transmutation in uranium fuel with a transuranic generation reduction fuel, which uses BWR fuel and increases the <sup>235</sup>U enrichment within the range of High-Assay Low-Enriched Uranium (HALEU). We evaluated the relationship between the total decay heat of <sup>241</sup>Am and <sup>244</sup>Cm. The increase in decay heat from <sup>244</sup>Cm when the <sup>235</sup>U enrichment is 3.8 wt%, which is the example case of LEU application as that of existing fuel. The generation of <sup>244</sup>Cm comes from a smaller mass number of TRU nuclides decrease as the neutron flux and neutron capture reactions decrease with increasing enrichment. As a result, in the range where the enrichment exceeds about 7.5 wt%, and the ratio of the decrease in the weight fraction of <sup>241</sup>Am to the increase in the weight fraction of <sup>244</sup>Cm decreases below 4.1 %, and the total decay heat of the two nuclides is lower than it would be in no-transmutation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111308"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143508813","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}

引用次数: 0

Study on CsI release rate coefficient from fuel for source term evaluation

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-02-27 DOI: 10.1016/j.anucene.2025.111288

Masami Taira , Takuma Fujiwara , Yuji Arita

Fission products released to the environment during severe accidents at nuclear power plants are evaluated using severe accident comprehensive analysis codes such as MELCOR. However, the codes cannot directly evaluate the release of CsI as molecular compound from fuel pellets. This is attributed that the release rate coefficient for molecular compounds such as CsI from fuel is not experimentally obtained yet. In this study, a method to experimentally derive the release rate coefficient of CsI from fuel was pursued. In the experiment, after simulated fuels containing CsI were fabricated using grain size as a parameter, CsI released from the simulated fuels was measured by a quadrupole mass spectrometer. Then the method to derive the release rate coefficients of CsI from fuel in the CORSOR-M model was developed converting experimental data into Arrhenius formula. The results demonstrated that the method can experimentally derive the molecular compounds release rate coefficients from fuel.

引用次数: 0

Adaptive learning observer based fixed-time stable controller for load following of a Pressurized Water Reactor

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-02-26 DOI: 10.1016/j.anucene.2025.111259

Qiming Xu , Hongliang Liu , Qizhen Xiao , Wenjie Zeng , Run Luo

Enhancing the load following capacity of Nuclear Power Plants (NPPs) remains a challenge. This paper focuses on designing a fixed-time stable controller for load following operation of a Pressurized Water Reactor (PWR) with compound disturbances. To compensate the limitation of current measuring techniques, an adaptive learning observer is firstly proposed to estimate the hard-to-measure states like Xenon concentration, Iodine concentration, average reactor fuel temperature and the compound disturbances of the PWR system. Based on these important information, in order to ensure that the output power of PWR can rapidly and accurately follow the prescribed idea power, a fixed-time stable controller is presented. Sequentially, Lyapunov method is employed to verify that the control system can be stable within a fixed-time, which upper bound can be accurately calculated by some parameters. Finally, simulation results are provided to illustrate the effectiveness of the designed controller in the load following operation and the performance of the adaptive learning observer.

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