Pub Date : 2026-01-21DOI: 10.1016/j.anucene.2026.112163
Yang Di , Zhang Chunyu , Lin Jiming , Ding Peng
Reduced order models (ROMs) have been widely adopted to accelerate high-fidelity simulations while retaining essential predictive accuracy. However, two gaps persist for POD-FV-ROMs in the context of thermal-hydraulic problems. First, when Dirichlet and Neumann boundaries coexist, a robust and effective strategy to enforce boundary constraints in the ROM is still missing. Second, the ROM performance has not been convincingly demonstrated on practical three-dimensional cases with complex geometries. To close these gaps, a ROM combining POD-Galerkin projection with supremizer stabilization and POD-RBF interpolation was investigated, together with a penalty formulation for boundary treatment. Two novel a priori estimators are proposed to determine penalty factors (PFs) without manual tuning, one based on residuals relevant to projected snapshots and one derived from optimization to a domain-wise error. The entire ROM framework was assessed on an 2 × 2 helical cruciform fuel assembly discretized into around 24 million cells with parameterized boundary conditions. The results demonstrated that, under appropriate PFs, the ROM delivered satisfying accuracy while achieving a speed up of five orders of magnitude.
{"title":"A tuning-free POD-FV-ROM with automatic boundary enforcement for practical thermal-hydraulic applications","authors":"Yang Di , Zhang Chunyu , Lin Jiming , Ding Peng","doi":"10.1016/j.anucene.2026.112163","DOIUrl":"10.1016/j.anucene.2026.112163","url":null,"abstract":"<div><div>Reduced order models (ROMs) have been widely adopted to accelerate high-fidelity simulations while retaining essential predictive accuracy. However, two gaps persist for POD-FV-ROMs in the context of thermal-hydraulic problems. First, when Dirichlet and Neumann boundaries coexist, a robust and effective strategy to enforce boundary constraints in the ROM is still missing. Second, the ROM performance has not been convincingly demonstrated on practical three-dimensional cases with complex geometries. To close these gaps, a ROM combining POD-Galerkin projection with supremizer stabilization and POD-RBF interpolation was investigated, together with a penalty formulation for boundary treatment. Two novel <em>a priori</em> estimators are proposed to determine penalty factors (PFs) without manual tuning, one based on residuals relevant to projected snapshots and one derived from optimization to a domain-wise error. The entire ROM framework was assessed on an 2 × 2 helical cruciform fuel assembly discretized into around 24 million cells with parameterized boundary conditions. The results demonstrated that, under appropriate PFs, the ROM delivered satisfying accuracy while achieving a speed up of five orders of magnitude.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112163"},"PeriodicalIF":2.3,"publicationDate":"2026-01-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035530","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}
Pub Date : 2026-01-21DOI: 10.1016/j.anucene.2026.112132
Restu Kojo
In risk-informed regulation, the significance of systems, structures, and components (SSCs) is assessed using multiple risk metrics, requiring a systematic method to determine whether SSC degradation has a greater impact on Level 1 or Level 2 probabilistic risk assessments (PRA). A key issue is that risk significance in Level 2 often exceeds that of Level 1 PRA due to the order-of-magnitude difference between target values for core damage frequency (CDF) and that of containment failure frequency (CFF). To address this, a new methodology was developed, including a novel measure—risk difference achievement worth (RDAW)—which enables transparent comparisons across different PRAs. This methodology was applied to large-scale PRA models and confirmed the consistency of the significance comparison results. In summary, a mathematically formulated methodology for comparing significance across multiple PRAs, which is applicable to large-scale practical models, has been established.
{"title":"Methodology for significance determination across multiple risk metrics using novel importance measures","authors":"Restu Kojo","doi":"10.1016/j.anucene.2026.112132","DOIUrl":"10.1016/j.anucene.2026.112132","url":null,"abstract":"<div><div>In risk-informed regulation, the significance of systems, structures, and components (SSCs) is assessed using multiple risk metrics, requiring a systematic method to determine whether SSC degradation has a greater impact on Level 1 or Level 2 probabilistic risk assessments (PRA). A key issue is that risk significance in Level 2 often exceeds that of Level 1 PRA due to the order-of-magnitude difference between target values for core damage frequency (CDF) and that of containment failure frequency (CFF). To address this, a new methodology was developed, including a novel measure—risk difference achievement worth (RDAW)—which enables transparent comparisons across different PRAs. This methodology was applied to large-scale PRA models and confirmed the consistency of the significance comparison results. In summary, a mathematically formulated methodology for comparing significance across multiple PRAs, which is applicable to large-scale practical models, has been established.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112132"},"PeriodicalIF":2.3,"publicationDate":"2026-01-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035531","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}
Pub Date : 2026-01-21DOI: 10.1016/j.anucene.2026.112157
Ruixiao Zhang , Yuhang Niu , Yanan He , Zhiwei Lu , Yingwei Wu , Jing Zhang , G.H. Su
Silicon carbide (SiC) composite cladding is a promising candidate for accident-tolerant fuel applications, exhibiting distinct thermomechanical and failure behaviors compared to metallic cladding such as Zircaloy under spent fuel storage conditions. Existing safety criteria for spent fuel storage may therefore be inapplicable to SiC-based assemblies. In this study, the updated FROBA code was used to simulate the performance of high-burnup SiC cladding spent fuel after reactor operation, considering long-term wet and dry storage as well as short-term off-normal dry storage. Results show that SiC cladding performs well during wet storage. During dry storage, the cladding stress slightly exceeds the 90 MPa reference limit for Zr cladding. Due to the probabilistic failure characteristics of monolithic SiC, this corresponds to an estimated failure probability of approximately 0.3%. Elevated internal rod pressure is the main contributor to this risk. The cladding peak temperature of 400℃ at the rod top indicates the highest failure risk. Lower storage temperature limits and optimized pressure balance can effectively mitigate failure.
{"title":"Performance assessment of spent fuel storage with duplex SiC composite cladding in Pressurized water reactors","authors":"Ruixiao Zhang , Yuhang Niu , Yanan He , Zhiwei Lu , Yingwei Wu , Jing Zhang , G.H. Su","doi":"10.1016/j.anucene.2026.112157","DOIUrl":"10.1016/j.anucene.2026.112157","url":null,"abstract":"<div><div>Silicon carbide (SiC) composite cladding is a promising candidate for accident-tolerant fuel applications, exhibiting distinct thermomechanical and failure behaviors compared to metallic cladding such as Zircaloy under spent fuel storage conditions. Existing safety criteria for spent fuel storage may therefore be inapplicable to SiC-based assemblies. In this study, the updated FROBA code was used to simulate the performance of high-burnup SiC cladding spent fuel after reactor operation, considering long-term wet and dry storage as well as short-term off-normal dry storage. Results show that SiC cladding performs well during wet storage. During dry storage, the cladding stress slightly exceeds the 90 MPa reference limit for Zr cladding. Due to the probabilistic failure characteristics of monolithic SiC, this corresponds to an estimated failure probability of approximately 0.3%. Elevated internal rod pressure is the main contributor to this risk. The cladding peak temperature of 400℃ at the rod top indicates the highest failure risk. Lower storage temperature limits and optimized pressure balance can effectively mitigate failure.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112157"},"PeriodicalIF":2.3,"publicationDate":"2026-01-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035532","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}
Many next generation reactors propose the use of nontraditional nuclear fuel in the form of pebbles filled with thousands of fuel particles. While these next generation pebble bed reactors have been in development for decades, recent support for nuclear energy has bolstered the process, with many reactor designs proposed for deployment in the coming decade. An issue facing pebble bed reactors is safeguarding the fuel itself. Research on the burnup measurement systems is still evolving for developing a nondestructive assay method to quantify the amount of fissile material present in a used fuel pebble, creating a challenge for international safeguards design. The study presented here investigates the potential of neutron self-interrogation of spent fuel pebbles as an innovative method to implement materials accountability in these advanced reactors. Through reactor physics and fuel burnup simulations, spent fuel pebble material compositions are found and a method is developed to equate the delayed gamma-ray emissions resulting from the fissions induced by the pebble neutrons to the mass of key fissile actinides. The feasibility of this self-interrogation method is assessed, leading to the conclusion that the method is suitable for use in a passive counting mode employing a 4π detection geometry. As an example, the mass of 235U, 238U, 239Pu and 241Pu can be predicted at a precision of 4.1%, 0.86%, 13% and 13%, respectively, when measuring 100 end-of-life spent fuel pebbles over approximately 12 days.
{"title":"Assessment of self-interrogation safeguards Signatures for pebble bed reactor fuel","authors":"Austen Ocanas , Farheen Naqvi , Sudeep Mitra , Jeremy Osborn","doi":"10.1016/j.anucene.2026.112155","DOIUrl":"10.1016/j.anucene.2026.112155","url":null,"abstract":"<div><div>Many next generation reactors propose the use of nontraditional nuclear fuel in the form of pebbles filled with thousands of fuel particles. While these next generation pebble bed reactors have been in development for decades, recent support for nuclear energy has bolstered the process, with many reactor designs proposed for deployment in the coming decade. An issue facing pebble bed reactors is safeguarding the fuel itself. Research on the burnup measurement systems is still evolving for developing a nondestructive assay method to quantify the amount of fissile material present in a used fuel pebble, creating a challenge for international safeguards design. The study presented here investigates the potential of neutron self-interrogation of spent fuel pebbles as an innovative method to implement materials accountability in these advanced reactors. Through reactor physics and fuel burnup simulations, spent fuel pebble material compositions are found and a method is developed to equate the delayed gamma-ray emissions resulting from the fissions induced by the pebble neutrons to the mass of key fissile actinides. The feasibility of this self-interrogation method is assessed, leading to the conclusion that the method is suitable for use in a passive counting mode employing a 4<em>π</em> detection geometry. As an example, the mass of <sup>235</sup>U, <sup>238</sup>U, <sup>239</sup>Pu and <sup>241</sup>Pu can be predicted at a precision of 4.1%, 0.86%, 13% and 13%, respectively, when measuring 100 end-of-life spent fuel pebbles over approximately 12 days.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112155"},"PeriodicalIF":2.3,"publicationDate":"2026-01-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035580","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}
This study introduces an improved non-dominated sorting genetic algorithm II (INSGA-II) to optimize the boundary delineation of sandstone-type uranium deposits by determining the threshold of uranium content per square meter (UCPSM, kg/m2) for minable units. A multi-objective optimization model was developed to maximize both economic and resource benefits, which was solved using the INSGA-II. Key enhancements include: (1) population initialization via symmetric Latin hypercube design (SLHD); (2) adaptive mutation and crossover parameters. Applied to real data from a Chinese mining area, the model and algorithm demonstrated practical effectiveness. The Pareto solution set derived from the optimization enabled the determination of UCPSM thresholds, supporting a novel mining-area boundary definition method based on aggregating minable units and unlocking the resource and potential economic value of idle uranium deposits. This approach offers a new decision-making tool for sandstone-type uranium mining area design.
{"title":"Techno-Economic optimization of sandstone uranium Mining: A Case study of uranium content per square meter","authors":"Jiabing Li , Chuanfei Zhang , Xiangxue Zhang , Meifang Chen , Mingtao Jia","doi":"10.1016/j.anucene.2026.112125","DOIUrl":"10.1016/j.anucene.2026.112125","url":null,"abstract":"<div><div>This study introduces an improved non-dominated sorting genetic algorithm II (INSGA-II) to optimize the boundary delineation of sandstone-type uranium deposits by determining the threshold of uranium content per square meter (UCPSM, kg/m<sup>2</sup>) for minable units. A multi-objective optimization model was developed to maximize both economic and resource benefits, which was solved using the INSGA-II. Key enhancements include: (1) population initialization via symmetric Latin hypercube design (SLHD); (2) adaptive mutation and crossover parameters. Applied to real data from a Chinese mining area, the model and algorithm demonstrated practical effectiveness. The Pareto solution set derived from the optimization enabled the determination of UCPSM thresholds, supporting a novel mining-area boundary definition method based on aggregating minable units and unlocking the resource and potential economic value of idle uranium deposits. This approach offers a new decision-making tool for sandstone-type uranium mining area design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112125"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035579","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}
Pub Date : 2026-01-20DOI: 10.1016/j.anucene.2026.112158
Xinling Dai , Dechang Cai , Jin Cai
Criticality safety analysis is essential for fuel assembly transport, as it ensures subcriticality under all potential accident scenarios. Traditional methods are computationally expensive, requiring hundreds of input cases. The Neutron Worth Iteration method was developed to efficiently determine conservative envelopes for PWR fuel assemblies under transport impacts. By iteratively adjusting the configuration of fuel rods according to the neutron worth distribution within a fuel assembly, the method optimizes rod positioning to maximize . This method significantly reduces computational effort while maintaining a slight deviation of less than 2.62. This approach enables rapid yet rigorous criticality safety assessments.
{"title":"Efficient critical safety analysis model for PWR fuel assembly under transport accidents by neutron worth iteration","authors":"Xinling Dai , Dechang Cai , Jin Cai","doi":"10.1016/j.anucene.2026.112158","DOIUrl":"10.1016/j.anucene.2026.112158","url":null,"abstract":"<div><div>Criticality safety analysis is essential for fuel assembly transport, as it ensures subcriticality under all potential accident scenarios. Traditional methods are computationally expensive, requiring hundreds of input cases. The Neutron Worth Iteration method was developed to efficiently determine conservative <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>∞</mi></mrow></msub></math></span> envelopes for PWR fuel assemblies under transport impacts. By iteratively adjusting the configuration of fuel rods according to the neutron worth distribution within a fuel assembly, the method optimizes rod positioning to maximize <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>∞</mi></mrow></msub></math></span>. This method significantly reduces computational effort while maintaining a slight <span><math><msub><mrow><mi>k</mi></mrow><mrow><mi>∞</mi></mrow></msub></math></span> deviation of less than 2.62<span><math><mo>‰</mo></math></span>. This approach enables rapid yet rigorous criticality safety assessments.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112158"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035575","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}
Pub Date : 2026-01-20DOI: 10.1016/j.anucene.2026.112141
H.C. Manjunatha , B.M. Sankarshan , P.S. Damodara Gupta , L. Seenappa , K.N. Sridhar , R. Munirathnam
This study investigates alternative materials to lead for radiation shielding, addressing the need for safer and more effective options. Traditional materials like lead, although effective due to their high atomic number, are toxic and pose environmental risks. The study explores a set of tantalum–tungsten–oxygen (Ta–W–O) compounds, including TaWO, TaWO, TaWO, TaWO, and others. These compounds offer promising shielding properties due to their high density, atomic number, and stability. Key shielding parameters such as mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), half-value layer (HVL), and effective atomic number (Z) were calculated and compared to lead. Among all the studied Ta–W–O compounds, TaWO3 was identified as the most efficient and thermodynamically stable lead-free shielding material, exhibiting the highest photon attenuation performance across low- and intermediate-energy ranges. Across various energy ranges, these compounds demonstrate superior radiation protection efficiency (RPE) and electron density, essential for shielding in healthcare, nuclear, and aerospace applications. The findings suggest that tantalum–tungsten compounds could serve as viable lead-free shielding materials, offering a safer and more sustainable alternative for radiation protection.
{"title":"Evaluation of tantalum–tungsten–oxygen compounds as lead-free radiation shielding materials","authors":"H.C. Manjunatha , B.M. Sankarshan , P.S. Damodara Gupta , L. Seenappa , K.N. Sridhar , R. Munirathnam","doi":"10.1016/j.anucene.2026.112141","DOIUrl":"10.1016/j.anucene.2026.112141","url":null,"abstract":"<div><div>This study investigates alternative materials to lead for radiation shielding, addressing the need for safer and more effective options. Traditional materials like lead, although effective due to their high atomic number, are toxic and pose environmental risks. The study explores a set of tantalum–tungsten–oxygen (Ta–W–O) compounds, including TaW<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<span><math><msub><mrow></mrow><mrow><mn>3</mn></mrow></msub></math></span>, TaW<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>, Ta<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>W<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<span><math><msub><mrow></mrow><mrow><mn>5</mn></mrow></msub></math></span>, TaWO<span><math><msub><mrow></mrow><mrow><mn>3</mn></mrow></msub></math></span>, and others. These compounds offer promising shielding properties due to their high density, atomic number, and stability. Key shielding parameters such as mass attenuation coefficient (MAC), linear attenuation coefficient (LAC), half-value layer (HVL), and effective atomic number (Z<span><math><msub><mrow></mrow><mrow><mtext>eff</mtext></mrow></msub></math></span>) were calculated and compared to lead. Among all the studied Ta–W–O compounds, TaW<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>O<sub>3</sub> was identified as the most efficient and thermodynamically stable lead-free shielding material, exhibiting the highest photon attenuation performance across low- and intermediate-energy ranges. Across various energy ranges, these compounds demonstrate superior radiation protection efficiency (RPE) and electron density, essential for shielding in healthcare, nuclear, and aerospace applications. The findings suggest that tantalum–tungsten compounds could serve as viable lead-free shielding materials, offering a safer and more sustainable alternative for radiation protection.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112141"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035577","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}
Pub Date : 2026-01-20DOI: 10.1016/j.anucene.2026.112153
Yilong Li , Yan Guo , Junyong Liu, Hongyu Yang, Guangyun Min, Naibin Jiang
This study aims to investigate the flow disturbance effect of different spacer grids in a Pressurized Water Reactor (PWR) and their influence on the flow-induced vibration characteristics of fuel rods. Based on the model from the Subchannel and Bundle Test (PSBT). Differences between two-way flow-structure interaction and one-way flow-structure interaction analyzed. Subsequently, the one-way flow-structure interaction method was adopted. Under the condition of consistent average inlet mean flow velocity, compared the results of each spacer grid subchannel: transverse flow, pressure distribution, and pressure drop. These effects show significant correlation with grid structural features such as rigid protrusion, spring, and mixing vanes. Thus amplitude are significant in subchannels: the influence of the simple supported grid is negligible, spacer grids with no mixing vanes has a significant influence, and spacer grids with mixing vanes have the most significant effects.
{"title":"Numerical investigation of spacer grid-induced flow disturbances and impact on fuel rod flow-induced vibrations","authors":"Yilong Li , Yan Guo , Junyong Liu, Hongyu Yang, Guangyun Min, Naibin Jiang","doi":"10.1016/j.anucene.2026.112153","DOIUrl":"10.1016/j.anucene.2026.112153","url":null,"abstract":"<div><div>This study aims to investigate the flow disturbance effect of different spacer grids in a Pressurized Water Reactor (PWR) and their influence on the flow-induced vibration characteristics of fuel rods. Based on the model from the Subchannel and Bundle Test (PSBT). Differences between two-way flow-structure interaction and one-way flow-structure interaction analyzed. Subsequently, the one-way flow-structure interaction method was adopted. Under the condition of consistent average inlet mean flow velocity, compared the results of each spacer grid subchannel: transverse flow, pressure distribution, and pressure drop. These effects show significant correlation with grid structural features such as rigid protrusion, spring, and mixing vanes. Thus amplitude are significant in subchannels: the influence of the simple supported grid is negligible, spacer grids with no mixing vanes has a significant influence, and spacer grids with mixing vanes have the most significant effects.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112153"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035578","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}
Pub Date : 2026-01-20DOI: 10.1016/j.anucene.2026.112147
Longji Qiu, Nan Chao, Yong-kuo Liu, Zongzhen Shi, Joseph Daniel
This paper proposes a gamma radiation field reconstruction method based on sparse detection data. By integrating compressed sensing (CS) theory with clustered transport theory, a sensing matrix for radiation field reconstruction is constructed. A novel Non-negative Lasso Orthogonal Matching Pursuit (NNLasso-OMP) algorithm is developed, combining the high efficiency and flexibility of OMP with the overfitting resistance of Lasso. To evaluate the proposed method, three simulation scenarios are conducted, using Monte Carlo simulation results as reference benchmarks. The reconstruction performance of NNLasso-OMP is compared with that of OMP, inverse distance weighting (IDW) and 3DCNN algorithms. Results show that the average relative error (ARE) of NNLasso-OMP remains below 10% across all scenarios, achieving a reconstruction success rate (SR) exceeding 95%, while accurately identifying source locations. The proposed NNLasso-OMP method outperforms both OMP and IDW, demonstrating its effectiveness in achieving high-quality gamma radiation field reconstruction from sparse measurements.
{"title":"A non-negative Lasso Orthogonal Matching Pursuit method for gamma radiation field reconstruction with sparse measurement data","authors":"Longji Qiu, Nan Chao, Yong-kuo Liu, Zongzhen Shi, Joseph Daniel","doi":"10.1016/j.anucene.2026.112147","DOIUrl":"10.1016/j.anucene.2026.112147","url":null,"abstract":"<div><div>This paper proposes a gamma radiation field reconstruction method based on sparse detection data. By integrating compressed sensing (CS) theory with clustered transport theory, a sensing matrix for radiation field reconstruction is constructed. A novel Non-negative Lasso Orthogonal Matching Pursuit (NNLasso-OMP) algorithm is developed, combining the high efficiency and flexibility of OMP with the overfitting resistance of Lasso. To evaluate the proposed method, three simulation scenarios are conducted, using Monte Carlo simulation results as reference benchmarks. The reconstruction performance of NNLasso-OMP is compared with that of OMP, inverse distance weighting (IDW) and 3DCNN algorithms. Results show that the average relative error (ARE) of NNLasso-OMP remains below 10% across all scenarios, achieving a reconstruction success rate (SR) exceeding 95%, while accurately identifying source locations. The proposed NNLasso-OMP method outperforms both OMP and IDW, demonstrating its effectiveness in achieving high-quality gamma radiation field reconstruction from sparse measurements.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112147"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035571","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}
Pub Date : 2026-01-20DOI: 10.1016/j.anucene.2025.112009
Restu Kojo
This study evaluates the thermal-hydraulics in the containment of boiling water reactors caused by superheated gas leakage under severe accident conditions. The research involved characterizing the heat transfer paths, which included identifying superheated gas leakage positions and heat release from the reactor pressure vessel boundaries, classifying the containment volumes, and categorizing potential failure sections of the containment. Based on these insights, the heat transfer paths during severe accidents were clarified, and accident scenarios considering leakage from the safety relief valve and traversing in-core probe tubes were selected as representative scenarios.
A three-dimensional computational fluid dynamics (CFD) model of the Mark I containment was developed to evaluate the thermal-hydraulics of an entire drywell of the containment. Special attention was given to modeling the detailed structures of the containment top head flange, radiation from the reactor pressure vessel upper head, condensation in the suppression pool, and heat release from the containment top head to the reactor well. The CFD analyses focused on two scenarios: safety relief valve leakage and traversing in-core probe tube leakage, which can result in significant temperature distribution in the upper and lower drywell, respectively. This study identified the high temperature location on the containment boundary with higher possibility of failure by high-temperature gas leakage under severe accident conditions based on temperature distribution obtained by the present detailed three-dimensional CFD analysis.
{"title":"Identification of the containment heating mechanism and temperature distribution by high-temperature gas leakage under severe accident conditions","authors":"Restu Kojo","doi":"10.1016/j.anucene.2025.112009","DOIUrl":"10.1016/j.anucene.2025.112009","url":null,"abstract":"<div><div>This study evaluates the thermal-hydraulics in the containment of boiling water reactors caused by superheated gas leakage under severe accident conditions. The research involved characterizing the heat transfer paths, which included identifying superheated gas leakage positions and heat release from the reactor pressure vessel boundaries, classifying the containment volumes, and categorizing potential failure sections of the containment. Based on these insights, the heat transfer paths during severe accidents were clarified, and accident scenarios considering leakage from the safety relief valve and traversing in-core probe tubes were selected as representative scenarios.</div><div>A three-dimensional computational fluid dynamics (CFD) model of the Mark I containment was developed to evaluate the thermal-hydraulics of an entire drywell of the containment. Special attention was given to modeling the detailed structures of the containment top head flange, radiation from the reactor pressure vessel upper head, condensation in the suppression pool, and heat release from the containment top head to the reactor well. The CFD analyses focused on two scenarios: safety relief valve leakage and traversing in-core probe tube leakage, which can result in significant temperature distribution in the upper and lower drywell, respectively. This study identified the high temperature location on the containment boundary with higher possibility of failure by high-temperature gas leakage under severe accident conditions based on temperature distribution obtained by the present detailed three-dimensional CFD analysis.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112009"},"PeriodicalIF":2.3,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035576","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}
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