Pub Date : 2026-08-01Epub Date: 2026-03-08DOI: 10.1016/j.anucene.2026.112227
Shuaiyu Xue , Chong Zhou , Hongxiang Yu , Pinyan Huang , Yang Zou
The passive residual heat removal system (PRHRS) enhances the inherent safety of molten salt reactors (MSRs) by enabling autonomous operation following an accident. However, its transient performance under accident scenarios requires further evaluation. This study applies a set of performance evaluation criteria for the PRHRS in a liquid-fueled MSR, focusing on key safety parameters such as temperature limits. Using the RELAP5-TMSR code, a transient system model was proposed to simulate design-basis accidents, including station blackout, primary pump seizure, and loss of flow in the secondary circuit. The results indicate that the PRHRS effectively maintains the fuel salt temperature below the 815 °C material limit while preventing solidification of the coolant salt. Furthermore, the system demonstrates a sufficient heat removal capacity to remove decay heat under extended accident conditions reliably. These findings demonstrate the effectiveness of the PRHRS for the studied reactor design and contribute to the safety assessment framework for liquid-fueled MSRs.
{"title":"Transient performance evaluation of passive residual heat removal system in liquid molten salt reactor","authors":"Shuaiyu Xue , Chong Zhou , Hongxiang Yu , Pinyan Huang , Yang Zou","doi":"10.1016/j.anucene.2026.112227","DOIUrl":"10.1016/j.anucene.2026.112227","url":null,"abstract":"<div><div>The passive residual heat removal system (PRHRS) enhances the inherent safety of molten salt reactors (MSRs) by enabling autonomous operation following an accident. However, its transient performance under accident scenarios requires further evaluation. This study applies a set of performance evaluation criteria for the PRHRS in a liquid-fueled MSR, focusing on key safety parameters such as temperature limits. Using the RELAP5-TMSR code, a transient system model was proposed to simulate design-basis accidents, including station blackout, primary pump seizure, and loss of flow in the secondary circuit. The results indicate that the PRHRS effectively maintains the fuel salt temperature below the 815 °C material limit while preventing solidification of the coolant salt. Furthermore, the system demonstrates a sufficient heat removal capacity to remove decay heat under extended accident conditions reliably. These findings demonstrate the effectiveness of the PRHRS for the studied reactor design and contribute to the safety assessment framework for liquid-fueled MSRs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112227"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387859","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 : 2026-08-01Epub Date: 2026-03-06DOI: 10.1016/j.anucene.2026.112266
Shlash A. Luhaib , Nassar Alnassar , Sultan J. Alsufyani , A. Saftah , Norah Alsairy , Mohamed Y.M. Mohsen , Sitah Alanazi , A. Abdelghafar Galahom
The global shortage of uranium, coupled with the increasing demand for clean energy and enhanced nuclear reactor safety, underscores the need to explore alternative accident-tolerant fuel (ATF) types. Thorium-carbon monoxide fuels are suggested to be used in the TRISO particle in one of the most promising Generation IV reactor, the small advanced high temperature reactor (SmAHTR). A three-dimensional SmAHTR core was modelled using the MCNPX code to examine its neutronic characteristics at the suggested fuels. The neutronic burn-up performance of the proposed fuels is analyzed during 1450 effective full power days (EFPDs). The effect of using Thorium-carbon monoxide fuels on the reactivity, actinides, non-actinides concentration and the reactivity temperature coefficient was examined. The radial power distribution through the core of the SmAHTR was analyzed to verify the feasibility of the suggested fuels. The results assured the potential advantage of using (Th-238U,233U-235U)CO and (Th-rgPu)CO (30 wt%) in SmAHTR reactor.
{"title":"Investigation of the feasibility of using thorium-carbon monoxide fuel as accident-tolerant fuel in small advanced high temperature reactor","authors":"Shlash A. Luhaib , Nassar Alnassar , Sultan J. Alsufyani , A. Saftah , Norah Alsairy , Mohamed Y.M. Mohsen , Sitah Alanazi , A. Abdelghafar Galahom","doi":"10.1016/j.anucene.2026.112266","DOIUrl":"10.1016/j.anucene.2026.112266","url":null,"abstract":"<div><div>The global shortage of uranium, coupled with the increasing demand for clean energy and enhanced nuclear reactor safety, underscores the need to explore alternative accident-tolerant fuel (ATF) types. Thorium-carbon monoxide fuels are suggested to be used in the TRISO particle in one of the most promising Generation IV reactor, the small advanced high temperature reactor (SmAHTR). A three-dimensional SmAHTR core was modelled using the MCNPX code to examine its neutronic characteristics at the suggested fuels. The neutronic burn-up performance of the proposed fuels is analyzed during 1450 effective full power days (EFPDs). The effect of using Thorium-carbon monoxide fuels on the reactivity, actinides, non-actinides concentration and the reactivity temperature coefficient was examined. The radial power distribution through the core of the SmAHTR was analyzed to verify the feasibility of the suggested fuels. The results assured the potential advantage of using (Th-<sup>238</sup>U,<sup>233</sup>U-<sup>235</sup>U)CO and (Th-rgPu)CO (30 wt%) in SmAHTR reactor.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112266"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387807","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 : 2026-08-01Epub Date: 2026-03-05DOI: 10.1016/j.anucene.2026.112223
Bassam A. Khuwaileh , Polina Matesha
This study presents NucBench, an open benchmark suite for assessing multimodal Large Language Models (LLMs) in nuclear engineering. NucBench includes curated quantitative, qualitative, and image-based tasks focused on pressurized and boiling water reactors, using standardized textual and visual inputs. As a community-driven resource, it supports consistent evaluation and development of reliable, domain-aware AI tools for reactor design and safety. Three advanced LLMs: GPT-4.1, Claude 3.7 Sonnet, and Gemini 2.5 Pro, were tested on undergraduate and reactor operator exams and two-phase flow image classification. GPT-4.1 achieved the highest quantitative accuracy (up to 95%) under deterministic settings, while Claude 3.7 Sonnet showed the greatest consistency in open-ended tasks. Performance declined and variability increased with higher sampling temperatures, especially for Gemini 2.5 Pro. Image classification accuracy peaked for Churn regimes but fell for complex flows. Results emphasize model tuning, explainability, and benchmark expansion for safety–critical applications.
{"title":"Establishing benchmarks for large language models in nuclear engineering: A NucBench evaluation of GPT-4.1, Claude 3.7 Sonnet, and Gemini 2.5 Pro","authors":"Bassam A. Khuwaileh , Polina Matesha","doi":"10.1016/j.anucene.2026.112223","DOIUrl":"10.1016/j.anucene.2026.112223","url":null,"abstract":"<div><div>This study presents NucBench, an open benchmark suite for assessing multimodal Large Language Models (LLMs) in nuclear engineering. NucBench includes curated quantitative, qualitative, and image-based tasks focused on pressurized and boiling water reactors, using standardized textual and visual inputs. As a community-driven resource, it supports consistent evaluation and development of reliable, domain-aware AI tools for reactor design and safety. Three advanced LLMs: GPT-4.1, Claude 3.7 Sonnet, and Gemini 2.5 Pro, were tested on undergraduate and reactor operator exams and two-phase flow image classification. GPT-4.1 achieved the highest quantitative accuracy (up to 95%) under deterministic settings, while Claude 3.7 Sonnet showed the greatest consistency in open-ended tasks. Performance declined and variability increased with higher sampling temperatures, especially for Gemini 2.5 Pro. Image classification accuracy peaked for Churn regimes but fell for complex flows. Results emphasize model tuning, explainability, and benchmark expansion for safety–critical applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112223"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387854","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 : 2026-08-01Epub Date: 2026-03-10DOI: 10.1016/j.anucene.2026.112268
Cong Liu , Shuang Tan , Junxia Wei
A Taylor basis discontinuous Galerkin finite element scheme is presented for the discrete ordinates (SN) transport equation in the two-dimensional r-z coordinate system. The basis functions based on the Taylor series expansion are hierarchical and independent of mesh shape, thus providing a unified framework for regular, deformed, and hybrid grids of triangles and quadrilaterals, which may be beneficial for some coupled simulation applications. The spatial distribution of the solution is represented by linear, bilinear, quadratic, and cubic polynomial expansion functions, respectively. Finite element integration is calculated in the physical element by Gaussian quadrature. By extending the basis function evaluation module, arbitrary higher-order DG schemes can be implemented. We develop a unified DG-SN code for the fixed-source, k-eigenvalue, α-eigenvalue, and time-dependent transport equations. Based on mesh priority and lagged angular flux, a sweep sorting algorithm is applied for decoupling dependency cycles between severely deformed grids. Numerical verification is performed for several transport problems, and the results indicate that our method achieves the theoretical convergence rate and strong robustness on non-orthogonal grids in r-z geometry. For the tested multi-medium transport problems, higher-order schemes exhibit certain computational advantages over lower-order schemes on coarse grids.
{"title":"Taylor basis DG scheme for SN transport equations on non-orthogonal grids in r-z coordinates","authors":"Cong Liu , Shuang Tan , Junxia Wei","doi":"10.1016/j.anucene.2026.112268","DOIUrl":"10.1016/j.anucene.2026.112268","url":null,"abstract":"<div><div>A Taylor basis discontinuous Galerkin finite element scheme is presented for the discrete ordinates (S<em><sub>N</sub></em>) transport equation in the two-dimensional r-z coordinate system. The basis functions based on the Taylor series expansion are hierarchical and independent of mesh shape, thus providing a unified framework for regular, deformed, and hybrid grids of triangles and quadrilaterals, which may be beneficial for some coupled simulation applications. The spatial distribution of the solution is represented by linear, bilinear, quadratic, and cubic polynomial expansion functions, respectively. Finite element integration is calculated in the physical element by Gaussian quadrature. By extending the basis function evaluation module, arbitrary higher-order DG schemes can be implemented. We develop a unified DG-S<sub>N</sub> code for the fixed-source, k-eigenvalue, α-eigenvalue, and time-dependent transport equations. Based on mesh priority and lagged angular flux, a sweep sorting algorithm is applied for decoupling dependency cycles between severely deformed grids. Numerical verification is performed for several transport problems, and the results indicate that our method achieves the theoretical convergence rate and strong robustness on non-orthogonal grids in r-z geometry. For the tested multi-medium transport problems, higher-order schemes exhibit certain computational advantages over lower-order schemes on coarse grids.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112268"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387812","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 : 2026-08-01Epub Date: 2026-03-07DOI: 10.1016/j.anucene.2026.112265
Reuven Rachamin , Astrid Barkleit , Jörg Konheiser , Marcus Seidl
The final shutdown of an NPP is followed by a post-operational phase, during which measures are taken to prepare the plant for decommissioning. One of the essential tasks in preparing the NPP for decommissioning is to obtain precise knowledge of the radioactivity content within the plant’s components, particularly in the RPV and its internal structures, which typically exhibit the highest levels of radioactivity. To address this challenge, a novel method combining two Monte Carlo codes, MCNP and FLUKA, was developed to evaluate the activation distribution within the components of an NPP. This paper provides an overview of the methodology and demonstrates its application through the activation calculations of selected RPV internal components of a 1300 MWe Siemens/KWU PWR. The calculations yielded results with high accuracy, demonstrating that the method can serve as a non-destructive tool for radiological characterization of the plant’s components.
{"title":"Radiological characterization of selected Siemens/KWU PWR components using the MCNP-FLUKA code sequence","authors":"Reuven Rachamin , Astrid Barkleit , Jörg Konheiser , Marcus Seidl","doi":"10.1016/j.anucene.2026.112265","DOIUrl":"10.1016/j.anucene.2026.112265","url":null,"abstract":"<div><div>The final shutdown of an NPP is followed by a post-operational phase, during which measures are taken to prepare the plant for decommissioning. One of the essential tasks in preparing the NPP for decommissioning is to obtain precise knowledge of the radioactivity content within the plant’s components, particularly in the RPV and its internal structures, which typically exhibit the highest levels of radioactivity. To address this challenge, a novel method combining two Monte Carlo codes, MCNP and FLUKA, was developed to evaluate the activation distribution within the components of an NPP. This paper provides an overview of the methodology and demonstrates its application through the activation calculations of selected RPV internal components of a 1300 MWe Siemens/KWU PWR. The calculations yielded results with high accuracy, demonstrating that the method can serve as a non-destructive tool for radiological characterization of the plant’s components.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112265"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387805","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 : 2026-08-01Epub Date: 2026-03-11DOI: 10.1016/j.anucene.2026.112267
Masahiko Nakase , Tatsuro Matsumura , Ryo Hamada , Chi Young Han , Go Chiba , Tomofumi Sakuragi , Hidekazu Asano
Japan is promoting a closed nuclear fuel cycle (NFC), the process of producing, using, and managing nuclear fuel from uranium mining to waste disposal, to enhance energy security and achieve a zero-carbon society. Since NFC involves multiple steps, an integrated approach is required to streamline the process, optimize conditions, and ultimately minimize the burden on geological disposal (GD). A key method identified is the separation of high-heat-generating minor actinides (MA), thereby reducing the size of the disposal repository. However, challenges such as optimal separation ratios and new fuel compositions require careful consideration. To address this, we employed a multi-criteria analysis (MCA) to assess various aspects of the NFC, dividing the process into GD, MA separation, and Fast Reactor (FR) core domains. Evaluation scores were assigned based on six criteria (CR), utilizing weighted sum and concordance analysis to determine optimal NFC conditions. One finding was that these evaluations correlate well, aiding in selecting technology options and setting research and development targets. The study shows that MCA effectively evaluates NFC options, revealing that optimal MA separation levels vary by criteria and confirming that simplified MA separation (not perfect MA separation) is a superior approach.
{"title":"Strategic selection of nuclear fuel cycle backend options: A framework incorporating MA separation and final disposal by multi-criteria evaluation","authors":"Masahiko Nakase , Tatsuro Matsumura , Ryo Hamada , Chi Young Han , Go Chiba , Tomofumi Sakuragi , Hidekazu Asano","doi":"10.1016/j.anucene.2026.112267","DOIUrl":"10.1016/j.anucene.2026.112267","url":null,"abstract":"<div><div>Japan is promoting a closed nuclear fuel cycle (NFC), the process of producing, using, and managing nuclear fuel from uranium mining to waste disposal, to enhance energy security and achieve a zero-carbon society. Since NFC involves multiple steps, an integrated approach is required to streamline the process, optimize conditions, and ultimately minimize the burden on geological disposal (GD). A key method identified is the separation of high-heat-generating minor actinides (MA), thereby reducing the size of the disposal repository. However, challenges such as optimal separation ratios and new fuel compositions require careful consideration. To address this, we employed a multi-criteria analysis (MCA) to assess various aspects of the NFC, dividing the process into GD, MA separation, and Fast Reactor (FR) core domains. Evaluation scores were assigned based on six criteria (CR), utilizing weighted sum and concordance analysis to determine optimal NFC conditions. One finding was that these evaluations correlate well, aiding in selecting technology options and setting research and development targets. The study shows that MCA effectively evaluates NFC options, revealing that optimal MA separation levels vary by criteria and confirming that simplified MA separation (not perfect MA separation) is a superior approach.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112267"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387810","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 : 2026-08-01Epub Date: 2026-03-10DOI: 10.1016/j.anucene.2026.112264
Hyun Jin Yu , Do Won Hyeon , Yong Hee Han , Sang Ryeol Yoon , Jong Soon Song , Min Ho Lee
Among large-scale metallic components, the steam generator (S/G) in particular, remains stored on-site due to the limited capacity of domestic disposal facilities. Accordingly, it is necessary to establish an effective dismantling and disposal strategy for the S/G. This study constructs a cutting and loading scenario for the S/G and quantitatively evaluates the radiological impact on workers. Cutting was performed along the weld lines and further segmented to match the dimensions of a 200 L drum, with the scenario designed to meet the required 85% drum fill rate. VISIPLAN and Integrated Modules for Bioassay Analysis (IMBA) codes were utilized to assess both external and internal radiation exposure. The external exposure dose was found to be 8.67 mSv/y for mechanical cutting and 1.34 mSv/y for thermal cutting. Internal exposure was evaluated as 0.33 mSv/y for Band Saw, 0.17 mSv/y for Reciprocating Saw, and 0.02 mSv/y for Plasma Arc. This study proposes a cutting and loading strategy that considers both radiation safety and waste volume reduction in decommissioning sites. The evaluation results are expected to contribute to the establishment of radiological regulatory standards and the development of safe dismantling strategies.
{"title":"Cutting and loading scenario and radiological dose assessment for spent steam generator decommissioning","authors":"Hyun Jin Yu , Do Won Hyeon , Yong Hee Han , Sang Ryeol Yoon , Jong Soon Song , Min Ho Lee","doi":"10.1016/j.anucene.2026.112264","DOIUrl":"10.1016/j.anucene.2026.112264","url":null,"abstract":"<div><div>Among large-scale metallic components, the steam generator (S/G) in particular, remains stored on-site due to the limited capacity of domestic disposal facilities. Accordingly, it is necessary to establish an effective dismantling and disposal strategy for the S/G. This study constructs a cutting and loading scenario for the S/G and quantitatively evaluates the radiological impact on workers. Cutting was performed along the weld lines and further segmented to match the dimensions of a 200 L drum, with the scenario designed to meet the required 85% drum fill rate. VISIPLAN and Integrated Modules for Bioassay Analysis (IMBA) codes were utilized to assess both external and internal radiation exposure. The external exposure dose was found to be 8.67 mSv/y for mechanical cutting and 1.34 mSv/y for thermal cutting. Internal exposure was evaluated as 0.33 mSv/y for Band Saw, 0.17 mSv/y for Reciprocating Saw, and 0.02 mSv/y for Plasma Arc. This study proposes a cutting and loading strategy that considers both radiation safety and waste volume reduction in decommissioning sites. The evaluation results are expected to contribute to the establishment of radiological regulatory standards and the development of safe dismantling strategies.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112264"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147387857","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 : 2026-08-01Epub Date: 2026-03-06DOI: 10.1016/j.anucene.2026.112241
Tom Drechsler , Sascha Weichel , Antonio Hurtado , Carsten Lange
In the context of Feynman- analysis, the bunching technique is a widely used method for synthesizing neutron count data with larger bin widths by aggregating counts from smaller bin widths. For each bin size , the variance-to-mean ratio is computed, forming the basis for determining the parameter. However, the points on the curve are inherently correlated due to the bunching process. As a result, uncorrelated fitting methods that rely solely on the standard errors of fail to provide accurate estimates for and its uncertainties. A proper treatment of these correlations requires incorporating the covariance matrix of the points into the fitting procedure. In practice, estimating this covariance matrix from real measurements is challenging and demands a large amount of data, while its theoretical estimation remains an open problem. This paper investigates alternative approaches to reliably determine and its uncertainties. Our analysis confirms that uncorrelated fits, neglecting the covariance matrix, fail to provide reliable uncertainties as correlations are significant. Conversely, including an accurately estimated covariance matrix yields correct results for and its uncertainties. Since direct estimation of the full covariance matrix requires extensive data, entailing significant measurement time and computational effort, a new method is proposed. This method enables the estimation of the necessary covariance information within practical limits of measurement time and computational resources. These findings reinforce the theoretical foundation of Feynman- analysis and offer a robust framework for accurately fitting correlated data arising from the bunching technique.
{"title":"A snippet-based algorithm for practical covariance estimation in Feynman-α analysis","authors":"Tom Drechsler , Sascha Weichel , Antonio Hurtado , Carsten Lange","doi":"10.1016/j.anucene.2026.112241","DOIUrl":"10.1016/j.anucene.2026.112241","url":null,"abstract":"<div><div>In the context of Feynman-<span><math><mi>α</mi></math></span> analysis, the bunching technique is a widely used method for synthesizing neutron count data with larger bin widths by aggregating counts from smaller bin widths. For each bin size <span><math><mi>T</mi></math></span>, the variance-to-mean ratio <span><math><mrow><mi>Y</mi><mrow><mo>(</mo><mi>T</mi><mo>)</mo></mrow></mrow></math></span> is computed, forming the basis for determining the <span><math><mi>α</mi></math></span> parameter. However, the points on the <span><math><mrow><mi>Y</mi><mrow><mo>(</mo><mi>T</mi><mo>)</mo></mrow></mrow></math></span> curve are inherently correlated due to the bunching process. As a result, uncorrelated fitting methods that rely solely on the standard errors of <span><math><mrow><mi>Y</mi><mrow><mo>(</mo><mi>T</mi><mo>)</mo></mrow></mrow></math></span> fail to provide accurate estimates for <span><math><mi>α</mi></math></span> and its uncertainties. A proper treatment of these correlations requires incorporating the covariance matrix of the <span><math><mrow><mi>Y</mi><mrow><mo>(</mo><mi>T</mi><mo>)</mo></mrow></mrow></math></span> points into the fitting procedure. In practice, estimating this covariance matrix from real measurements is challenging and demands a large amount of data, while its theoretical estimation remains an open problem. This paper investigates alternative approaches to reliably determine <span><math><mi>α</mi></math></span> and its uncertainties. Our analysis confirms that uncorrelated fits, neglecting the covariance matrix, fail to provide reliable uncertainties as correlations are significant. Conversely, including an accurately estimated covariance matrix yields correct results for <span><math><mi>α</mi></math></span> and its uncertainties. Since direct estimation of the full covariance matrix requires extensive data, entailing significant measurement time and computational effort, a new method is proposed. This method enables the estimation of the necessary covariance information within practical limits of measurement time and computational resources. These findings reinforce the theoretical foundation of Feynman-<span><math><mi>α</mi></math></span> analysis and offer a robust framework for accurately fitting correlated data arising from the bunching technique.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"233 ","pages":"Article 112241"},"PeriodicalIF":2.3,"publicationDate":"2026-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"147388277","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 : 2026-06-01Epub Date: 2026-01-09DOI: 10.1016/j.anucene.2026.112122
Jingxiang Cao , Guifeng Zhu , Huiqin Yin , Linbing Jiang , Xinmei Yang , Jie Qiu , Wenguan Liu
Tellurium (Te) is the primary cause of intergranular embrittlement in structural materials of molten salt reactors (MSR). This study investigates the chemical behavior of Te in FLiBe molten salt under different redox states, and finds that the redox states have a substantial impact on the chemical behavior of Te. Under reducing redox conditions of the molten salt, Te can stably exist as an anion and preferentially forms bonds with positively charged Th or U atoms. In neutral or mildly oxidizing environments, Te atoms are more likely to aggregate and form Te–Te bonds, which facilitates nucleation in the molten salt and promotes their adsorption onto the alloy surface. Under strongly oxidizing conditions, Te tends to exist in a cationic form and may be present as tellurium fluoride gas. This study reveals the possibility of inhibiting Te-induced intergranular embrittlement in MSRs by adjusting the redox state of the molten salt.
{"title":"Effect of molten salt redox states on the chemical behavior of Tellurium: A machine learning molecular dynamics study","authors":"Jingxiang Cao , Guifeng Zhu , Huiqin Yin , Linbing Jiang , Xinmei Yang , Jie Qiu , Wenguan Liu","doi":"10.1016/j.anucene.2026.112122","DOIUrl":"10.1016/j.anucene.2026.112122","url":null,"abstract":"<div><div>Tellurium (Te) is the primary cause of intergranular embrittlement in structural materials of molten salt reactors (MSR). This study investigates the chemical behavior of Te in FLiBe molten salt under different redox states, and finds that the redox states have a substantial impact on the chemical behavior of Te. Under reducing redox conditions of the molten salt, Te can stably exist as an anion and preferentially forms bonds with positively charged Th or U atoms. In neutral or mildly oxidizing environments, Te atoms are more likely to aggregate and form Te–Te bonds, which facilitates nucleation in the molten salt and promotes their adsorption onto the alloy surface. Under strongly oxidizing conditions, Te tends to exist in a cationic form and may be present as tellurium fluoride gas. This study reveals the possibility of inhibiting Te-induced intergranular embrittlement in MSRs by adjusting the redox state of the molten salt.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112122"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145915165","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 : 2026-06-01Epub Date: 2026-01-22DOI: 10.1016/j.anucene.2026.112143
Qifan Wang , Minyun Liu , Yanping Huang , Shanfang Huang , Ruohan Zheng , Houjun Gong
This study employs the lattice Boltzmann method to investigate the dynamics and heat transfer of carbon dioxide droplets on wall surfaces. Droplet spreading is shown to follow three distinct stages—rupture-to-rest, inertia–viscosity transition, and viscosity-dominated—each governed by power-law scaling of contact radius with time, with the rupture-to-rest stage unique to static droplets. Simulations across varying impact heights and wall inclinations demonstrate the universality of this scaling, extending classical spreading laws beyond flat surfaces. Droplet thickness evolution proceeds through accelerated shrinkage, decelerated shrinkage, and eventual stabilization. For inclined walls, the contact-line center emerges as a key dynamical reference, enabling systematic characterization of asymmetric spreading and sliding. Heat transfer analysis further identifies the cryogenic ring phenomenon as a dominant mechanism of wall conduction, while wall superheating and wettability are shown to strongly modulate thermal performance. These findings establish a unified framework for droplet morphology and heat transfer under realistic conditions.
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