Pub Date : 2024-11-13DOI: 10.1016/j.anucene.2024.111030
Deepak K. Solanki , Kausik Nandi , Joe Mohan , Arunkumar Sridharan , S.V. Prabhu
Subcooled flow boiling of water is widely observed in high heat flux and high mass flux (HHHM) cooling applications such as heat exchangers, refrigeration equipment, boiler tubes and nuclear reactor core fuel channels in pressurized heavy water reactors (PHWR). In this study, the focus is on investigating the local heat transfer coefficient (HTC) and pressure drop in a horizontal tube experiencing subcooled boiling of water under low pressure and HHHM conditions. The study encompasses different geometrical parameters such as tube diameter (5.5 mm, 7.5 mm, 9.5 mm and 12 mm) and length (550 mm for each of the tubes). The operating parameters that are varied include mass flux (248–2000 kg/m2.s) and heat flux (0–1837 kW/m2). Infrared thermography is used to measure the local wall temperature. A non-dimensional correlation for the diabatic pressure drop ratio (ratio of diabatic pressure drop to adiabatic pressure drop) as a function of Jakob number (), Boiling number () and diameter ratio is developed. Subcooled boiling pressure drop ratio for 5.5 mm, 7.5 mm and 9.4 mm diameter tubes is 2.23 which is independent of diameter. A correlation for the two phase local HTC during subcooled flow boiling conditions as a function of , and Prandtl number () is also developed.
{"title":"Subcooled flow boiling in a horizontal circular pipe under high heat flux and high mass flux conditions","authors":"Deepak K. Solanki , Kausik Nandi , Joe Mohan , Arunkumar Sridharan , S.V. Prabhu","doi":"10.1016/j.anucene.2024.111030","DOIUrl":"10.1016/j.anucene.2024.111030","url":null,"abstract":"<div><div>Subcooled flow boiling of water is widely observed in high heat flux and high mass flux (HHHM) cooling applications such as heat exchangers, refrigeration equipment, boiler tubes and nuclear reactor core fuel channels in pressurized heavy water reactors (PHWR). In this study, the focus is on investigating the local heat transfer coefficient (HTC) and pressure drop in a horizontal tube experiencing subcooled boiling of water under low pressure and HHHM conditions. The study encompasses different geometrical parameters such as tube diameter (5.5 <em>mm</em>, 7.5 <em>mm</em>, 9.5 <em>mm</em> and 12 <em>mm</em>) and length (550 <em>mm</em> for each of the tubes). The operating parameters that are varied include mass flux (248–2000 <em>kg/m</em><sup><em>2</em></sup><sup><em>.</em></sup><em>s</em>) and heat flux (0–1837 <em>kW/m</em><sup><em>2</em></sup>). Infrared thermography is used to measure the local wall temperature. A non-dimensional correlation for the diabatic pressure drop ratio (ratio of diabatic pressure drop to adiabatic pressure drop) as a function of Jakob number (<span><math><mrow><mi>Ja</mi></mrow></math></span>), Boiling number (<span><math><mrow><mi>Bo</mi></mrow></math></span>) and diameter ratio is developed. Subcooled boiling pressure drop ratio for 5.5 <em>mm</em>, 7.5 <em>mm</em> and 9.4 <em>mm</em> diameter tubes is 2.23 which is independent of diameter. A correlation for the two phase local HTC during subcooled flow boiling conditions as a function of <span><math><mrow><mi>Ja</mi></mrow></math></span>, <span><math><mrow><mi>Bo</mi></mrow></math></span> and Prandtl number (<span><math><mrow><mi>Pr</mi></mrow></math></span>) is also developed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111030"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658377","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 : 2024-11-13DOI: 10.1016/j.anucene.2024.111045
Yeongjun Kim , Haneol Lee , Man-Sung Yim
The current nuclear safeguards approach to spent nuclear fuel inspection at nuclear power stations is based on item counting and limited partial defect analysis. With the expected surge in spent fuel storage, limited spent fuel storage pool capacity, and the increasing need for transferring fuel to long-term storage facilities, there is a growing demand for more efficient and cost-effective nuclear safeguards approaches for nuclear materials management in civilian nuclear power facilities. This study proposes a scintillator-based indirect gamma detector for spent fuel inventory screening inspection, specifically designed for use in interim storage pools prior to fuel transfer to difficult-to-access storage. This paper presents the design of the proposed detector, its application for spent fuel screening inventory inspection, and analysis using MCNP for partial defect detection. Results of analysis indicated that verifying a ∼ 13.6 % level of randomly distributed fuel defect for the Westinghouse 17x17 fuel assembly is possible using this approach. The performance evaluation also indicates that inspection of spent fuel assemblies of various vendor types against 1 SQ diversion may be possible.
{"title":"Investigation of fast and cost-effective partial defect detector for spent fuel transfer verification to enhance nuclear safeguards","authors":"Yeongjun Kim , Haneol Lee , Man-Sung Yim","doi":"10.1016/j.anucene.2024.111045","DOIUrl":"10.1016/j.anucene.2024.111045","url":null,"abstract":"<div><div>The current nuclear safeguards approach to spent nuclear fuel inspection at nuclear power stations is based on item counting and limited partial defect analysis. With the expected surge in spent fuel storage, limited spent fuel storage pool capacity, and the increasing need for transferring fuel to long-term storage facilities, there is a growing demand for more efficient and cost-effective nuclear safeguards approaches for nuclear materials management in civilian nuclear power facilities. This study proposes a scintillator-based indirect gamma detector for spent fuel inventory screening inspection, specifically designed for use in interim storage pools prior to fuel transfer to difficult-to-access storage. This paper presents the design of the proposed detector, its application for spent fuel screening inventory inspection, and analysis using MCNP for partial defect detection. Results of analysis indicated that verifying a ∼ 13.6 % level of randomly distributed fuel defect for the Westinghouse 17x17 fuel assembly is possible using this approach. The performance evaluation also indicates that inspection of spent fuel assemblies of various vendor types against 1 SQ diversion may be possible.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111045"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658372","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}
According to the design uncertainty limit of reactor physics response calculation, target accuracy assessment solves a inequality-nonlinear-constrained optimization problem and provides the priorities of nuclear data uncertainty requirements, which is beneficial to the improvement of safety and economy of nuclear reactors. Due to the ultra-large amount of nuclear data, solving the optimization problem in full-space is improbable. Subspace method could reduce calculation space dimensions effectively and improve the stability of numerical solution, in the meantime the high-dimensional information is mostly retained. Studies on nuclear data target accuracy assessment for the first core of China Experimental Fast Reactor (CEFR) at BOC and EOC based on subspace method shows that, within the 0.3% target accuracy of effective multiplication factor, computational dimension can be reduced from over 1000 to less than 100. The most affected reactions are the scattering and radiation capture of 23Na, 52Cr, 56Fe and 235U. The results of this paper provide a comparable target accuracy assessment for similar fast reactors utilizing enriched uranium dioxide fuel, and are helpful for communicating across pipeline from evaluator to end user.
{"title":"Target accuracy assessment for China Experimental Fast Reactor based on subspace method","authors":"Yaxin Qiao , Xiaofei Wu , Ping Liu , Haicheng Wu , Huanyu Zhang , Lili Wen , Ying Chen , Yue Xiao","doi":"10.1016/j.anucene.2024.111022","DOIUrl":"10.1016/j.anucene.2024.111022","url":null,"abstract":"<div><div>According to the design uncertainty limit of reactor physics response calculation, target accuracy assessment solves a inequality-nonlinear-constrained optimization problem and provides the priorities of nuclear data uncertainty requirements, which is beneficial to the improvement of safety and economy of nuclear reactors. Due to the ultra-large amount of nuclear data, solving the optimization problem in full-space is improbable. Subspace method could reduce calculation space dimensions effectively and improve the stability of numerical solution, in the meantime the high-dimensional information is mostly retained. Studies on nuclear data target accuracy assessment for the first core of China Experimental Fast Reactor (CEFR) at BOC and EOC based on subspace method shows that, within the 0.3% target accuracy of effective multiplication factor, computational dimension can be reduced from over 1000 to less than 100. The most affected reactions are the scattering and radiation capture of <sup>23</sup>Na, <sup>52</sup>Cr, <sup>56</sup>Fe and <sup>235</sup>U. The results of this paper provide a comparable target accuracy assessment for similar fast reactors utilizing enriched uranium dioxide fuel, and are helpful for communicating across pipeline from evaluator to end user.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111022"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658376","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}
Tungsten granular blockage accident is one of standard accidents in the CiADS system. In this paper, the benchmark experiments for the tungsten granular blockage accident have been performed on the VENUS-II light water zero-power facility of CIAE in Beijing. The effective multiplication factor Keff values of the reactor were measured by the source jerk method for the tungsten granular targets with different heights, 5 cm, 10 cm, 20 cm, 30 cm and 40 cm. To achieve the proper sub-critical status, 904 fuel rods were loaded in the reactor core. The results were analyzed with the Monte Carlo code NECP-MCX using the ENDF/B-VII.0 library. In the NECP-MCX simulations, the efficient homogeneous modeling method was adopted to build the tungsten granular target model. The trend of the simulated Keff values was found to be similar with those of the experiment during tungsten granular blockage accident, and the minimum and maximum relative deviations are respectively about 0.38 % and 0.51 %. The results show that the effective multiplication factor Keff decrease with the increase of tungsten particle height and the Keff curves are generally inverse S-shaped. The sensitivity analyses in NECP-MCX indicate that the negative reactivity worth of tungsten granular target is mainly caused by the radiation capture reaction channel in the tungsten alloy grains. Through further analyses, it was found that the tungsten (n, gamma) reaction channel in the thermal and epithermal neutron regions is most related to the Keff variation of the VENUS-II light water zero-power facility. Meanwhile, the tungsten granular target has the great influence on the neutron flux and neutron spectra around the spallation target zone. To improve the security of the CiADS integration system, the efficient method to avoid the tungsten granular target blockage accident should be recommended.
{"title":"Investigation on neutronics changes in tungsten granular blockage accident on VENUS-II light water reactor","authors":"Liang Chen , Wei Jiang , Zhao-Dong Xia , Fei Ma , Rui Yu , Hai-Yan Meng , Qi Zhou , Hong-Lin Ge , Yan-Bin Zhang , Xue-Ying Zhang","doi":"10.1016/j.anucene.2024.111038","DOIUrl":"10.1016/j.anucene.2024.111038","url":null,"abstract":"<div><div>Tungsten granular blockage accident is one of standard accidents in the CiADS system. In this paper, the benchmark experiments for the tungsten granular blockage accident have been performed on the VENUS-II light water zero-power facility of CIAE in Beijing. The effective multiplication factor K<sub>eff</sub> values of the reactor were measured by the source jerk method for the tungsten granular targets with<!--> <!-->different<!--> <!-->heights, 5 cm, 10 cm, 20 cm, 30 cm and 40 cm. To achieve the proper sub-critical status, 904 fuel rods were loaded in the reactor core. The results were analyzed with the Monte Carlo code NECP-MCX using the ENDF/B-VII.0 library. In the NECP-MCX simulations, the efficient homogeneous modeling method was adopted to build the tungsten granular target model. The<!--> <!-->trend<!--> <!-->of the simulated K<sub>eff</sub> values was<!--> <!-->found to be similar<!--> <!-->with<!--> <!-->those of the experiment during tungsten granular blockage accident, and the minimum and maximum relative deviations are respectively about 0.38 % and 0.51 %. The results show that the effective multiplication factor K<sub>eff</sub> decrease with the increase of tungsten particle height and the K<sub>eff</sub> curves are generally inverse S-shaped. The sensitivity analyses in NECP-MCX indicate that the negative reactivity worth of tungsten granular target is mainly caused by the radiation capture reaction channel in the tungsten alloy grains. Through further analyses, it was found that the tungsten (n, gamma) reaction channel in the thermal and epithermal neutron regions is most related to the K<sub>eff</sub> variation of the VENUS-II light water zero-power facility. Meanwhile, the tungsten granular target has the great influence on the neutron flux and neutron spectra around the spallation target zone. To improve the<!--> <!-->security<!--> <!-->of<!--> <!-->the<!--> <!-->CiADS integration system, the efficient method to avoid the tungsten granular target blockage accident should be recommended.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111038"},"PeriodicalIF":1.9,"publicationDate":"2024-11-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658374","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 : 2024-10-29DOI: 10.1016/j.anucene.2024.111010
Zaiyong Ma , Qi Wu , Yugao Ma , Simiao Tang , Luteng Zhang , Changwen Liu , Liangming Pan
Many studies have been carried out on the geyser boiling phenomenon in liquid metal heat pipes, but most of them are phenomenological. Considering that nucleation should be the trigger of geyser boiling, occurrence of geyser boiling should be predicted by studying related nucleation characteristics. In this paper, the experimental value of nucleation superheat were gained and analyzed. The results showed that the main parameters affecting the nucleation superheat of sodium heat pipe were heating power and pressure, and the influences of wick mesh number, inclination angle, liquid filling ratio etc. were not important. By fitting the experimental data with heat flux and wall cavity vapor pressure, a correlation which could predict the data well was proposed. Further investigation showed that there may be materials with bad wetting properties in the wall cavities, whose surface thermo-physical properties could be greatly affected by temperature, resulting in the decrease of nucleation contact angle with increase in wall temperature.
{"title":"The nucleation characteristics of geyser boiling in sodium heat pipes","authors":"Zaiyong Ma , Qi Wu , Yugao Ma , Simiao Tang , Luteng Zhang , Changwen Liu , Liangming Pan","doi":"10.1016/j.anucene.2024.111010","DOIUrl":"10.1016/j.anucene.2024.111010","url":null,"abstract":"<div><div>Many studies have been carried out on the geyser boiling phenomenon in liquid metal heat pipes, but most of them are phenomenological. Considering that nucleation should be the trigger of geyser boiling, occurrence of geyser boiling should be predicted by studying related nucleation characteristics. In this paper, the experimental value of nucleation superheat were gained and analyzed. The results showed that the main parameters affecting the nucleation superheat of sodium heat pipe were heating power and pressure, and the influences of wick mesh number, inclination angle, liquid filling ratio etc. were not important. By fitting the experimental data with heat flux and wall cavity vapor pressure, a correlation which could predict the data well was proposed. Further investigation showed that there may be materials with bad wetting properties in the wall cavities, whose surface thermo-physical properties could be greatly affected by temperature, resulting in the decrease of nucleation contact angle with increase in wall temperature.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111010"},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142526953","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 : 2024-10-29DOI: 10.1016/j.anucene.2024.110963
Anže Pungerčič , Vicente Bécares , Daniel Cano-Ott , Roberta Cirillo , Tom Clarijs , Jacek Gajewski , Bor Kos , Renata Mikołajczak , Evžen Novák , Gabriel Pavel , Georg Pohlner , Lisanne Van Puyvelde , Jörg Starflinger , László Szentmiklósi , Joanna Walkiewicz , Luka Snoj
Nuclear research reactors ( RR ) are essential facilities in countries implementing nuclear power plants and are used for experiments necessary for commercial reactor development, training and education programs, and many other applications not related to nuclear energy production (e.g., isotope production, neutron sources, materials science). Europe has a broad and very diverse landscape of RRs, many of which have been in operation for 30-60 years, are well maintained and regularly modernized. However, financial pressures caused by a combination of declining interest and the lack of a sound financial model have led to the closure of many of them (e.g. OSIRIS in Saclay, JEEP II research reactor at IFE Kjeller and BER2 in Berlin). These negative trends called for coordinated European action to assess the impact of the declining number of RRs. The Towards Optimized Use of Research Reactors (TOURR) project was a response to this challenge. Its main objective was to assess the status of the EU RR fleet and to develop a strategy for the refurbishment and construction of new RR in Europe. The assessment was based on analysed data obtained through extensive questionnaires sent to all operating European RR. The analysis revealed gaps in terms of lack of long-term funding, lack of manpower and lack of communication between RRs and their customers. It also showed threats of further European RR closures. Regarding the long-term EU RR strategy, the main recommendations of the TOURR project are to build (at least) two RRs, a medium-flux multipurpose reactor and a flexible zero-power facility. Both reactor cores could be part of a single facility built at the European level and accessible to all EU Member States.
{"title":"European research reactor strategy derived in the scope of the towards optimized use of research reactors (TOURR) project","authors":"Anže Pungerčič , Vicente Bécares , Daniel Cano-Ott , Roberta Cirillo , Tom Clarijs , Jacek Gajewski , Bor Kos , Renata Mikołajczak , Evžen Novák , Gabriel Pavel , Georg Pohlner , Lisanne Van Puyvelde , Jörg Starflinger , László Szentmiklósi , Joanna Walkiewicz , Luka Snoj","doi":"10.1016/j.anucene.2024.110963","DOIUrl":"10.1016/j.anucene.2024.110963","url":null,"abstract":"<div><div>Nuclear research reactors ( RR ) are essential facilities in countries implementing nuclear power plants and are used for experiments necessary for commercial reactor development, training and education programs, and many other applications not related to nuclear energy production (e.g., isotope production, neutron sources, materials science). Europe has a broad and very diverse landscape of RRs, many of which have been in operation for 30-60 years, are well maintained and regularly modernized. However, financial pressures caused by a combination of declining interest and the lack of a sound financial model have led to the closure of many of them (e.g. OSIRIS in Saclay, JEEP II research reactor at IFE Kjeller and BER2 in Berlin). These negative trends called for coordinated European action to assess the impact of the declining number of RRs. The Towards Optimized Use of Research Reactors (TOURR) project was a response to this challenge. Its main objective was to assess the status of the EU RR fleet and to develop a strategy for the refurbishment and construction of new RR in Europe. The assessment was based on analysed data obtained through extensive questionnaires sent to all operating European RR. The analysis revealed gaps in terms of lack of long-term funding, lack of manpower and lack of communication between RRs and their customers. It also showed threats of further European RR closures. Regarding the long-term EU RR strategy, the main recommendations of the TOURR project are to build (at least) two RRs, a medium-flux multipurpose reactor and a flexible zero-power facility. Both reactor cores could be part of a single facility built at the European level and accessible to all EU Member States.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110963"},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527009","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}
Pub Date : 2024-10-29DOI: 10.1016/j.anucene.2024.111011
Long Yun, Xu Yuan, Guo Xi’an, Zhang Mingyu
This paper presents a study on the internal flow characteristics of reactor coolant pumps using Dynamic Mode Decomposition (DMD) technology. As a core component of nuclear power plants, the internal flow characteristics of reactor coolant pumps play a crucial role in the performance and stability of the pumps. This paper initially introduces the application of DMD and Proper Orthogonal Decomposition (POD) methods in fluid mechanics, emphasizing the effectiveness of DMD in analyzing the dynamic characteristics of flow fields. A computational model of the reactor coolant pump was constructed, and numerical simulation of the internal flow field under non-uniform inflow conditions was conducted. The impact of the lower chamber of the steam generator on the pump’s inlet conditions was evaluated. The numerical simulation results were analyzed using DMD technology, extracting flow characteristics and revealing the main flow modes and dynamic behaviors in the flow field. The results demonstrate that the DMD technology can accurately capture the time-dynamic characteristics within the flow field, providing crucial insights for optimizing performance and preventing faults in the reactor coolant pump.
{"title":"Analysis of internal flow excitation characteristics of reactor coolant pump based on DMD","authors":"Long Yun, Xu Yuan, Guo Xi’an, Zhang Mingyu","doi":"10.1016/j.anucene.2024.111011","DOIUrl":"10.1016/j.anucene.2024.111011","url":null,"abstract":"<div><div>This paper presents a study on the internal flow characteristics of reactor coolant pumps using Dynamic Mode Decomposition (DMD) technology. As a core component of nuclear power plants, the internal flow characteristics of reactor coolant pumps play a crucial role in the performance and stability of the pumps. This paper initially introduces the application of DMD and Proper Orthogonal Decomposition (POD) methods in fluid mechanics, emphasizing the effectiveness of DMD in analyzing the dynamic characteristics of flow fields. A computational model of the reactor coolant pump was constructed, and numerical simulation of the internal flow field under non-uniform inflow conditions was conducted. The impact of the lower chamber of the steam generator on the pump’s inlet conditions was evaluated. The numerical simulation results were analyzed using DMD technology, extracting flow characteristics and revealing the main flow modes and dynamic behaviors in the flow field. The results demonstrate that the DMD technology can accurately capture the time-dynamic characteristics within the flow field, providing crucial insights for optimizing performance and preventing faults in the reactor coolant pump.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111011"},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527010","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 : 2024-10-29DOI: 10.1016/j.anucene.2024.110975
Đ. Petrović , A. Rineiski , M. Zanetti , G. Scheveneels , X.-N. Chen , William D’haeseleer
Accidental scenarios that involve degradation of a fast reactor core and/or relocation of its fuel material call for particular attention with respect to promptcritical reactivity events. The dynamics of a power transient during such an event is governed by the Prompt Neutron Generation Time (PNGT), a parameter that is sensitive to the moderating power of the system. Since Heavy Liquid Metals (HLMs) have a boiling point that is higher than the melting point of stainless steel, the first degradation mechanism to occur in a Heavy Liquid Metal Fast Reactor (HLMFR) is likely to be the loss of structural material. This sequence holds the potential to create conditions for considerable spectrum softening and may thus have important implications for the value of the PNGT. In the framework of this study, a (postulated) complete absence of structural material in and around the reactor core of an HLMFR is demonstrated to lead to an increase in the PNGT by an order of magnitude.
The sensitivity of the fission energy release during a promptcritical event to the value of the PNGT is further investigated by employing the severe accident code SIMMER-III. To correctly model a degraded reactor core characterized by a considerably softer neutron spectrum when compared to the spectrum of its intact configuration, a new neutron data set is generated. This is done by introducing new energy groups to the already existing 11-energy-group structure and collapsing multigroup cross-section data by employing a weighting spectrum representative of a degraded core configuration of an HLMFR. Subsequent simulations demonstrate that an increase in the PNGT by a factor of ∼4 yields an increase in the fission energy release during a Core Disruptive Accident (CDA) by ∼50 %. It is therefore established that low-energy neutrons may play an important role during a promptcritical reactivity transient in an HLMFR.
{"title":"On the Neutron Kinetics during a Promptcritical Accident in a Heavy Liquid Metal Fast Reactor and the Importance of Low-Energy Neutrons","authors":"Đ. Petrović , A. Rineiski , M. Zanetti , G. Scheveneels , X.-N. Chen , William D’haeseleer","doi":"10.1016/j.anucene.2024.110975","DOIUrl":"10.1016/j.anucene.2024.110975","url":null,"abstract":"<div><div>Accidental scenarios that involve degradation of a fast reactor core and/or relocation of its fuel material call for particular attention with respect to promptcritical reactivity events. The dynamics of a power transient during such an event is governed by the Prompt Neutron Generation Time (PNGT), a parameter that is sensitive to the moderating power of the system. Since Heavy Liquid Metals (HLMs) have a boiling point that is higher than the melting point of stainless steel, the first degradation mechanism to occur in a Heavy Liquid Metal Fast Reactor (HLMFR) is likely to be the loss of structural material. This sequence holds the potential to create conditions for considerable spectrum softening and may thus have important implications for the value of the PNGT. In the framework of this study, a (postulated) complete absence of structural material in and around the reactor core of an HLMFR is demonstrated to lead to an increase in the PNGT by an order of magnitude.</div><div>The sensitivity of the fission energy release during a promptcritical event to the value of the PNGT is further investigated by employing the severe accident code SIMMER-III. To correctly model a degraded reactor core characterized by a considerably softer neutron spectrum when compared to the spectrum of its intact configuration, a new neutron data set is generated. This is done by introducing new energy groups to the already existing 11-energy-group structure and collapsing multigroup cross-section data by employing a weighting spectrum representative of a degraded core configuration of an HLMFR. Subsequent simulations demonstrate that an increase in the PNGT by a factor of ∼4 yields an increase in the fission energy release during a Core Disruptive Accident (CDA) by ∼50 %. It is therefore established that low-energy neutrons may play an important role during a promptcritical reactivity transient in an HLMFR.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110975"},"PeriodicalIF":1.9,"publicationDate":"2024-10-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527014","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 : 2024-10-28DOI: 10.1016/j.anucene.2024.111003
Yuan Chang , Hui Wang , Gong-Lin Li , Ming Ding
An in-house code called CONTHAC-3D was developed to investigate the fundamental thermal–hydraulic phenomena occurred in the containment under severe accidents for NPPs. The code included specific models to simulate the special systems of HPR1000 and ACP100. The classical backward-facing step flow benchmark and BMC HYJET helium jet experiments were selected to investigate the code’s capability of simulating hydrogen diffusion process. The results showed that the difference between the calculated and experimental results could be negligible. The code was then applied to investigate hydrogen diffusion and distribution for HPR1000. The results showed that the hydrogen released from the break rises vertically and rapidly to the containment dome, then the gas diffused into the dome and lower compartments. As the time went by, the hydrogen concentration in lower compartments seemed to be higher than that in the containment dome. The results could provide foundation for the arrangement of hydrogen risk mitigation measures.
{"title":"Research on simulation of hydrogen diffusion behavior based on CONTHAC-3D code","authors":"Yuan Chang , Hui Wang , Gong-Lin Li , Ming Ding","doi":"10.1016/j.anucene.2024.111003","DOIUrl":"10.1016/j.anucene.2024.111003","url":null,"abstract":"<div><div>An in-house code called CONTHAC-3D was developed to investigate the fundamental thermal–hydraulic phenomena occurred in the containment under severe accidents for NPPs. The code included specific models to simulate the special systems of HPR1000 and ACP100. The classical backward-facing step flow benchmark and BMC HYJET helium jet experiments were selected to investigate the code’s capability of simulating hydrogen diffusion process. The results showed that the difference between the calculated and experimental results could be negligible. The code was then applied to investigate hydrogen diffusion and distribution for HPR1000. The results showed that the hydrogen released from the break rises vertically and rapidly to the containment dome, then the gas diffused into the dome and lower compartments. As the time went by, the hydrogen concentration in lower compartments seemed to be higher than that in the containment dome. The results could provide foundation for the arrangement of hydrogen risk mitigation measures.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 111003"},"PeriodicalIF":1.9,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527102","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 : 2024-10-28DOI: 10.1016/j.anucene.2024.110978
J.L. Wormald, A.J. Trainer, M.L. Zerkle
A new method is presented to use machine-learned interatomic potentials (MLPs) to generate material models for thermal neutron scattering laws (TSLs). MLPs are computationally efficient models of ab initio force fields that can be used in the creation of a vibrational spectrum as an input to TSL generation. MLP-based molecular dynamics introduces temperature effects into the vibrational spectrum, which have been neglected in most modern TSLs. Yttrium hydride () is used to illustrate this new MLP technique. The MLP approach is shown to predict temperature effects in the vibrational spectrum observed in experiment and improve on key features of the oscillatory scattering cross section of when compared to current temperature-independent, ab initio techniques.
{"title":"Machine-learned force fields for thermal neutron scattering law evaluations","authors":"J.L. Wormald, A.J. Trainer, M.L. Zerkle","doi":"10.1016/j.anucene.2024.110978","DOIUrl":"10.1016/j.anucene.2024.110978","url":null,"abstract":"<div><div>A new method is presented to use machine-learned interatomic potentials (MLPs) to generate material models for thermal neutron scattering laws (TSLs). MLPs are computationally efficient models of <em>ab initio</em> force fields that can be used in the creation of a vibrational spectrum as an input to TSL generation. MLP-based molecular dynamics introduces temperature effects into the vibrational spectrum, which have been neglected in most modern TSLs. Yttrium hydride (<span><math><msub><mrow><mi>YH</mi></mrow><mrow><mi>x</mi></mrow></msub></math></span>) is used to illustrate this new MLP technique. The MLP approach is shown to predict temperature effects in the vibrational spectrum observed in experiment and improve on key features of the oscillatory scattering cross section of <span><math><msub><mrow><mi>YH</mi></mrow><mrow><mi>x</mi></mrow></msub></math></span> when compared to current temperature-independent, <em>ab initio</em> techniques.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"211 ","pages":"Article 110978"},"PeriodicalIF":1.9,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142527101","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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