Pub Date : 2025-02-18DOI: 10.1016/j.nucengdes.2025.113921
Yanlin Li, Benke Qin, Hanliang Bo, Wen He
The real-time information of the control rods position should be provided by the rod position indicators. Therefore, the rod position indicators (RPIs) are the pivotal safety-related equipment of the nuclear reactors. According to the operating properties of the 200 MW nuclear reactor (NHR200-Ⅱ), the measuring channel of the RPIs fills with two-phase mediums, as the nitrogen and the water. The liquid level of the measuring channel is affected by the operating temperature of the NHR200-Ⅱ. Based on the coaxial type capacitance RPIs, the impact of the rod position and the liquid level on the output of the indicator are studied separately. The position solving model of the control rods is proposed. Based on the model, the dual-channel capacitance RPIs is designed, and the precision of the indicator is evaluated under the cold state and the ordinary pressure. Results indicate that the maximum error of the indicator is within ±4 mm.
{"title":"A control rods position recognition method based capacitance sensing principles within varying liquid level environments","authors":"Yanlin Li, Benke Qin, Hanliang Bo, Wen He","doi":"10.1016/j.nucengdes.2025.113921","DOIUrl":"10.1016/j.nucengdes.2025.113921","url":null,"abstract":"<div><div>The real-time information of the control rods position should be provided by the rod position indicators. Therefore, the rod position indicators (RPIs) are the pivotal safety-related equipment of the nuclear reactors. According to the operating properties of the 200 MW nuclear reactor (NHR200-Ⅱ), the measuring channel of the RPIs fills with two-phase mediums, as the nitrogen and the water. The liquid level of the measuring channel is affected by the operating temperature of the NHR200-Ⅱ. Based on the coaxial type capacitance RPIs, the impact of the rod position and the liquid level on the output of the indicator are studied separately. The position solving model of the control rods is proposed. Based on the model, the dual-channel capacitance RPIs is designed, and the precision of the indicator is evaluated under the cold state and the ordinary pressure. Results indicate that the maximum error of the indicator is within ±4 mm.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113921"},"PeriodicalIF":1.9,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429686","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 : 2025-02-18DOI: 10.1016/j.nucengdes.2025.113917
Liang Yao , Rui-Tao Liu , Dan-Dan Su , Su-Ming Wang , Lu Wang , Hong-Na Zhang , Xiao-Bin Li , Feng-Chen Li
In space nuclear power systems, heat exchangers account for more than 50% of the total system weight. To minimize both weight and volume, this study selected a compact plate-fin heat exchanger (PFHE) as the NaK78-He heat exchanger for the space nuclear power system. Computational fluid dynamics (CFD) simulations were employed to analyze the velocity distribution at the NaK78 inlet side of a typical PFHE structure, and the channel design was optimized to enhance heat transfer and reduce pressure loss. To investigate the performance variations of the PFHE and further decrease its weight and volume, a simplified and accurate quasi-three-dimensional (Q3D) numerical calculation method was proposed. This method was combined with a genetic algorithm to optimize the PFHE’s size parameters. Results indicated that the uneven flow distribution on the NaK78 side could reduce the heat flow by approximately 10% and increase pressure loss by over 500%. The optimized PFHE significantly improved flow uniformity, reducing the standard deviation of velocity distribution from 1.11 to 0.26. Furthermore, after optimization using the genetic algorithm, the PFHE achieved a 23.47% reduction in overall volume and a 22.38% decrease in weight. This study provides valuable insights for optimizing heat exchangers in space nuclear power systems and presents an efficient, rapid, and accurate method for performance evaluation and optimization in engineering applications.
{"title":"Performance analysis and structural optimization of NaK78-He plate-fin heat exchanger for space nuclear reactor power systems based on a Q3D numerical method","authors":"Liang Yao , Rui-Tao Liu , Dan-Dan Su , Su-Ming Wang , Lu Wang , Hong-Na Zhang , Xiao-Bin Li , Feng-Chen Li","doi":"10.1016/j.nucengdes.2025.113917","DOIUrl":"10.1016/j.nucengdes.2025.113917","url":null,"abstract":"<div><div>In space nuclear power systems, heat exchangers account for more than 50% of the total system weight. To minimize both weight and volume, this study selected a compact plate-fin heat exchanger (PFHE) as the NaK<sup>78</sup>-He heat exchanger for the space nuclear power system. Computational fluid dynamics (CFD) simulations were employed to analyze the velocity distribution at the NaK<sup>78</sup> inlet side of a typical PFHE structure, and the channel design was optimized to enhance heat transfer and reduce pressure loss. To investigate the performance variations of the PFHE and further decrease its weight and volume, a simplified and accurate quasi-three-dimensional (Q3D) numerical calculation method was proposed. This method was combined with a genetic algorithm to optimize the PFHE’s size parameters. Results indicated that the uneven flow distribution on the NaK<sup>78</sup> side could reduce the heat flow by approximately 10% and increase pressure loss by over 500%. The optimized PFHE significantly improved flow uniformity, reducing the standard deviation of velocity distribution from 1.11 to 0.26. Furthermore, after optimization using the genetic algorithm, the PFHE achieved a 23.47% reduction in overall volume and a 22.38% decrease in weight. This study provides valuable insights for optimizing heat exchangers in space nuclear power systems and presents an efficient, rapid, and accurate method for performance evaluation and optimization in engineering applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113917"},"PeriodicalIF":1.9,"publicationDate":"2025-02-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429688","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 : 2025-02-17DOI: 10.1016/j.nucengdes.2025.113910
M. Venkatesh , S.Sathish Kumar , G. Parthiban , V. Snehalatha , M. Bootharajan , M. Muthuganesh , E. Prabhu , P.K. Chaurasia , V.Suresh Kumar , R. Sudha , E.Hemanth Rao , K. Sundararajan , Sanjay Kumar Das , Rajesh Ganesan , V. Jayaraman , D. Ponraju , N. Sivaraman
This study investigates the interaction between liquid sodium and cellulose wool at various temperatures, commonly used in fast breeder reactors (FBR). Cellulose wool, used for cleaning reactor components, was subjected to sodium interaction experiments within the welded stainless steel capsules in argon heated at temperatures from 150 °C to 550 °C. Analysis was carried out for the gaseous reaction products and solid residue using various characterization techniques. The products showed the presence of H2, CO, CO2, CH4, C2H2, C2H4, C2H6 and residual carbon. Gaseous products from the reactions were analyzed using Gas Chromatography (GC) with flame ionisation detector (GC-FID), Discharge ionization detector (GC-DID) and metal oxide semiconducting sensors. The gaseous products at 150 °C indicated the presence of CO and CO2 as the sample temperature was increased these gases were absent. The concentration of H2, C2H4 and C2H6 in the gaseous product decreased while the CH4 concentration increased with temperature. Solid residues were analysed by (i) Scanning electron microscope/Energy dispersive (SEM/EDS) for morphology and chemical content, (ii) carbon analysis and (iii) infrared spectroscopy (IR). The systematic study using these techniques revealed a significant decomposition of cellulose wool with liquid sodium at high temperatures.
{"title":"Studies on the interaction of sodium with cellulose wool","authors":"M. Venkatesh , S.Sathish Kumar , G. Parthiban , V. Snehalatha , M. Bootharajan , M. Muthuganesh , E. Prabhu , P.K. Chaurasia , V.Suresh Kumar , R. Sudha , E.Hemanth Rao , K. Sundararajan , Sanjay Kumar Das , Rajesh Ganesan , V. Jayaraman , D. Ponraju , N. Sivaraman","doi":"10.1016/j.nucengdes.2025.113910","DOIUrl":"10.1016/j.nucengdes.2025.113910","url":null,"abstract":"<div><div>This study investigates the interaction between liquid sodium and cellulose wool at various temperatures, commonly used in fast breeder reactors (FBR). Cellulose wool, used for cleaning reactor components, was subjected to sodium interaction experiments within the welded stainless steel capsules in argon heated at temperatures from 150 °C to 550 °C. Analysis was carried out for the gaseous reaction products and solid residue using various characterization techniques. The products showed the presence of H<sub>2</sub>, CO, CO<sub>2</sub>, CH<sub>4</sub>, C<sub>2</sub>H<sub>2</sub>, C<sub>2</sub>H<sub>4</sub>, C<sub>2</sub>H<sub>6</sub> and residual carbon. Gaseous products from the reactions were analyzed using Gas Chromatography (GC) with flame ionisation detector (GC-FID), Discharge ionization detector (GC-DID) and metal oxide semiconducting sensors. The gaseous products at 150 °C indicated the presence of CO and CO<sub>2</sub> as the sample temperature was increased these gases were absent. The concentration of H<sub>2</sub>, C<sub>2</sub>H<sub>4</sub> and C<sub>2</sub>H<sub>6</sub> in the gaseous product decreased while the CH<sub>4</sub> concentration increased with temperature. Solid residues were analysed by (i) Scanning electron microscope/Energy dispersive (SEM/EDS) for morphology and chemical content, (ii) carbon analysis and (iii) infrared spectroscopy (IR). The systematic study using these techniques revealed a significant decomposition of cellulose wool with liquid sodium at high temperatures.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113910"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429684","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 : 2025-02-17DOI: 10.1016/j.nucengdes.2025.113914
Ming Liu , Pan Wu , Ke Huang , Jianqiang Shan
Space nuclear reactor power system can offer high energy density and a wide power coverage, providing electric power for space exploration. The safety characteristics of space nuclear reactor during in-orbit operation accidents is key to evaluate its feasibility. In this paper, a safety analysis code is developed for He-Xe mixture cooled space reactor based on SCTRAN, which used to be applied for Supercritical water or CO2 cooled reactor. A semi empirical formula is applied to develop a physical property calculation model of He–Xe. Flow resistance and heat transfer correlations as well as turbomachinery models for He-Xe are developed. A numerical model for S4-CBC space reactor power system is established by using the newly developed code. The calculation results of reactor and system model have the maximum relative errors of 3.43 % and 0.47 % compared to the reference design values, which proves the code is reliable to simulate the thermal–hydraulic behavior of the S4-CBC space reactor power system. The transient responses of the S4-CBC space reactor power system under the reactivity insertion accident and the mechanical failure of one and two Brayton loops are simulated and analyzed. The results of the accident analysis provide guidance for the core design, operation strategy, and safety system design of future He-Xe cooled space reactor system.
{"title":"Development and application of safety analysis code for He-Xe cooled space reactor","authors":"Ming Liu , Pan Wu , Ke Huang , Jianqiang Shan","doi":"10.1016/j.nucengdes.2025.113914","DOIUrl":"10.1016/j.nucengdes.2025.113914","url":null,"abstract":"<div><div>Space nuclear reactor power system can offer high energy density and a wide power coverage, providing electric power for space exploration. The safety characteristics of space nuclear reactor during in-orbit operation accidents is key to evaluate its feasibility. In this paper, a safety analysis code is developed for He-Xe mixture cooled space reactor based on SCTRAN, which used to be applied for Supercritical water or CO<sub>2</sub> cooled reactor. A semi empirical formula is applied to develop a physical property calculation model of He–Xe. Flow resistance and heat transfer correlations as well as turbomachinery models for He-Xe are developed. A numerical model for S<sup>4</sup>-CBC space reactor power system is established by using the newly developed code. The calculation results of reactor and system model have the maximum relative errors of 3.43 % and 0.47 % compared to the reference design values, which proves the code is reliable to simulate the thermal–hydraulic behavior of the S<sup>4</sup>-CBC space reactor power system. The transient responses of the S<sup>4</sup>-CBC space reactor power system under the reactivity insertion accident and the mechanical failure of one and two Brayton loops are simulated and analyzed. The results of the accident analysis provide guidance for the core design, operation strategy, and safety system design of future He-Xe cooled space reactor system.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113914"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429685","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 : 2025-02-17DOI: 10.1016/j.nucengdes.2025.113897
Pengfei Gao , Yanhui Li , Zhouyang Bai , Shaoming Ding , Yinan Zhang , Limei Xing , Zhihong Yu
Stress corrosion cracking (SCC) which has long been an essential topic for alloys exposed to subcritical water systems,is an urgent issue for pressurized water reactors (PWRs). Due to the complexity of the interaction of materials, environment, and stress, the SCC mechanism of consensus and the prediction model with mechanic extensibility are unavailable. Based on reviewing the typical characteristics and critical influence factors of SCC of stainless steels, this paper systematically summarizes, compares the more mainstream cracking mechanisms, and from the perspective of empirical and deterministic methods, discusses the applicable conditions of a series of SCC prediction models with high recognition via practical data, which are of great significance for strengthening readers’ understanding of SCC critical parameters and the matching service environment of each SCC mechanism, promoting the accurate prediction of SCC behavior and cumulative damage in the pressurized water reactors.
{"title":"Behavior mechanism and prediction models of stress corrosion cracking of stainless steels in pressurized water reactors","authors":"Pengfei Gao , Yanhui Li , Zhouyang Bai , Shaoming Ding , Yinan Zhang , Limei Xing , Zhihong Yu","doi":"10.1016/j.nucengdes.2025.113897","DOIUrl":"10.1016/j.nucengdes.2025.113897","url":null,"abstract":"<div><div>Stress corrosion cracking (SCC) which has long been an essential topic for alloys exposed to subcritical water systems,is an urgent issue for pressurized water reactors (PWRs). Due to the complexity of the interaction of materials, environment, and stress, the SCC mechanism of consensus and the prediction model with mechanic extensibility are unavailable. Based on reviewing the typical characteristics and critical influence factors of SCC of stainless steels, this paper systematically summarizes, compares the more mainstream cracking mechanisms, and from the perspective of empirical and deterministic methods, discusses the applicable conditions of a series of SCC prediction models with high recognition via practical data, which are of great significance for strengthening readers’ understanding of SCC critical parameters and the matching service environment of each SCC mechanism, promoting the accurate prediction of SCC behavior and cumulative damage in the pressurized water reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113897"},"PeriodicalIF":1.9,"publicationDate":"2025-02-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143429689","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 : 2025-02-16DOI: 10.1016/j.nucengdes.2025.113913
Shuhua Zhou, Ruizhi Hao, Wenshu Jiang, Shuo Wang, Hanwu Gao, Xinkui Fang, Kuo Bian, Xue Chen, Tao Lu
During the Steam Generator Tube Rupture (SGTR) accident of Lead-cooled Fast Reactors (LFR), liquid metals come into direct contact with water, resulting in a multiphase flow phenomenon involving liquid metal, water and vapor. Studying the corresponding mechanisms of the interaction process between subcooled water jet and liquid metal is crucial with the pressure impact and temperature drastic changes during the phenomenon may damage the structural components of the reactors. The simulation experiments were conducted to investigate the interaction characteristics of subcooled water jet with high-pressure upward into high-temperature thermal oil in a confined space, studying the effect of experimental parameters (such as injection time and pressure, oil and water temperature, nozzle shape and diameter) on the oil temperature drop and pressure variation. Three pressurization modes were distinguished, including two overall vapor productions and one local vapor production. The evolution path of water/vapor in oil was inferred based on transient temperature changes. Finally, the oil temperature drop model was established by four dimensionless thermophysical parameters, and the model was validated with experimental values at an average error of 12.05%.
{"title":"Interaction characteristics of subcooled water jet with high-pressure upward into high-temperature oil in a confined space","authors":"Shuhua Zhou, Ruizhi Hao, Wenshu Jiang, Shuo Wang, Hanwu Gao, Xinkui Fang, Kuo Bian, Xue Chen, Tao Lu","doi":"10.1016/j.nucengdes.2025.113913","DOIUrl":"10.1016/j.nucengdes.2025.113913","url":null,"abstract":"<div><div>During the Steam Generator Tube Rupture (SGTR) accident of Lead-cooled Fast Reactors (LFR), liquid metals come into direct contact with water, resulting in a multiphase flow phenomenon involving liquid metal, water and vapor. Studying the corresponding mechanisms of the interaction process between subcooled water jet and liquid metal is crucial with the pressure impact and temperature drastic changes during the phenomenon may damage the structural components of the reactors. The simulation experiments were conducted to investigate the interaction characteristics of subcooled water jet with high-pressure upward into high-temperature thermal oil in a confined space, studying the effect of experimental parameters (such as injection time and pressure, oil and water temperature, nozzle shape and diameter) on the oil temperature drop and pressure variation. Three pressurization modes were distinguished, including two overall vapor productions and one local vapor production. The evolution path of water/vapor in oil was inferred based on transient temperature changes. Finally, the oil temperature drop model was established by four dimensionless thermophysical parameters, and the model was validated with experimental values at an average error of 12.05%.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113913"},"PeriodicalIF":1.9,"publicationDate":"2025-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143421987","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 : 2025-02-16DOI: 10.1016/j.nucengdes.2025.113909
Mingda Han , Hetong Liu , Weixu Zhang , Yibo Zhang , Songhui Luo
The primary method for assessing cladding failure due to internal gas pressure involves comparing peak and ultimate pressures. When a small amount of gas is present, the pressure decreases as the volume expands at high temperatures, allowing the cladding to operate safely. This paper proposes a gas equilibrium-based failure criterion for cladding, determined by comparing the sectional force required by the gas with what the structure can provide. Finite element analysis compared cladding failure under three different boundary conditions (most used), using bidirectional fluid–structure interaction to investigate failure from internal gas. The critical load-bearing gas moles and the cross-sectional force changes with radial displacement for both hollow cladding and cladding with a fuel pellet were investigated. It was found that the critical gas quantity ratio is significantly smaller than the volume ratio, with gas pressure decreasing more rapidly in the pellet case.
{"title":"A failure criterion for nuclear fuel cladding due to internal gas","authors":"Mingda Han , Hetong Liu , Weixu Zhang , Yibo Zhang , Songhui Luo","doi":"10.1016/j.nucengdes.2025.113909","DOIUrl":"10.1016/j.nucengdes.2025.113909","url":null,"abstract":"<div><div>The primary method for assessing cladding failure due to internal gas pressure involves comparing peak and ultimate pressures. When a small amount of gas is present, the pressure decreases as the volume expands at high temperatures, allowing the cladding to operate safely. This paper proposes a gas equilibrium-based failure criterion for cladding, determined by comparing the sectional force required by the gas with what the structure can provide. Finite element analysis compared cladding failure under three different boundary conditions (most used), using bidirectional fluid–structure interaction to investigate failure from internal gas. The critical load-bearing gas moles and the cross-sectional force changes with radial displacement for both hollow cladding and cladding with a fuel pellet were investigated. It was found that the critical gas quantity ratio is significantly smaller than the volume ratio, with gas pressure decreasing more rapidly in the pellet case.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113909"},"PeriodicalIF":1.9,"publicationDate":"2025-02-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143421988","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 : 2025-02-14DOI: 10.1016/j.nucengdes.2025.113916
Vesa Riikonen, Virpi Kouhia, Markku Puustinen, Giteshkumar Patel, Antti Räsänen, Eetu Kotro, Juhani Hyvärinen
The presence of non-condensable gases in the reactor cooling system can significantly influence the operation of several safety systems within a nuclear power plant. The pressurized nitrogen volume at the top of the accumulator tank is the driving force for injecting accumulator water. In certain nuclear power plants, the release of gaseous nitrogen to the primary side is inhibited by an automatic closure of the accumulator injection line at the end of the discharge. However, if this automatic closure system fails, nitrogen will inadvertently flow into the reactor cooling system once the accumulators have been depleted. Furthermore, it is important also to note that the water within the accumulator is saturated with dissolved nitrogen, resulting in the injection of some nitrogen into the primary system alongside every accumulator discharge.
The impact of nitrogen on core cooling during loss of coolant accidents (LOCAs) has been investigated experimentally using the PWR PACTEL facility. The main observations were that when a break occurs in the hot leg, the injection of nitrogen from an accumulator can effectively prevent depressurization of the primary side. Consequently, the core can experience a heat-up at primary pressure somewhat above the typical low-pressure safety injection (LPSI) shut-off head. The decoupling of primary and secondary side pressures depends on the amount of nitrogen released from the accumulator, how much of it accumulates into the U-shaped steam generator heat exchange tubes, decreasing the condensation heat transfer, and the number of steam generators participating in the secondary side depressurization. Furthermore, the size of the break significantly affects the volume of nitrogen escaping through the break, which in turn influences the nitrogen levels within the system. Specifically, larger breaks permit a greater flow of nitrogen, thereby reducing the likelihood of disrupting heat transfer between the primary and secondary sides while also mitigating the depressurization of the primary side.
{"title":"Experimental studies on nitrogen’s effect on reactor core cooling during a hot leg SBLOCA in a scaled EPR model","authors":"Vesa Riikonen, Virpi Kouhia, Markku Puustinen, Giteshkumar Patel, Antti Räsänen, Eetu Kotro, Juhani Hyvärinen","doi":"10.1016/j.nucengdes.2025.113916","DOIUrl":"10.1016/j.nucengdes.2025.113916","url":null,"abstract":"<div><div>The presence of non-condensable gases in the reactor cooling system can significantly influence the operation of several safety systems within a nuclear power plant. The pressurized nitrogen volume at the top of the accumulator tank is the driving force for injecting accumulator water. In certain nuclear power plants, the release of gaseous nitrogen to the primary side is inhibited by an automatic closure of the accumulator injection line at the end of the discharge. However, if this automatic closure system fails, nitrogen will inadvertently flow into the reactor cooling system once the accumulators have been depleted. Furthermore, it is important also to note that the water within the accumulator is saturated with dissolved nitrogen, resulting in the injection of some nitrogen into the primary system alongside every accumulator discharge.</div><div>The impact of nitrogen on core cooling during loss of coolant accidents (LOCAs) has been investigated experimentally using the PWR PACTEL facility. The main observations were that when a break occurs in the hot leg, the injection of nitrogen from an accumulator can effectively prevent depressurization of the primary side. Consequently, the core can experience a heat-up at primary pressure somewhat above the typical low-pressure safety injection (LPSI) shut-off head. The decoupling of primary and secondary side pressures depends on the amount of nitrogen released from the accumulator, how much of it accumulates into the U-shaped steam generator heat exchange tubes, decreasing the condensation heat transfer, and the number of steam generators participating in the secondary side depressurization. Furthermore, the size of the break significantly affects the volume of nitrogen escaping through the break, which in turn influences the nitrogen levels within the system. Specifically, larger breaks permit a greater flow of nitrogen, thereby reducing the likelihood of disrupting heat transfer between the primary and secondary sides while also mitigating the depressurization of the primary side.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113916"},"PeriodicalIF":1.9,"publicationDate":"2025-02-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143421985","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 : 2025-02-14DOI: 10.1016/j.nucengdes.2025.113911
Xingwei Chen , Zhizhe Xu , Dai Ye , Yang Zou
The heat pipe-cooled micro molten salt reactor (micro-MSR) utilizes heat pipes for transferring the fission energy produced in the core to thermoelectric generators (TEG). In order to assess the heat transfer performance, an integrated experimental setup comprising a molten salt – heat pipe – thermoelectric generator was established. Experiments were carried out to assess the system’s performance during start-up and operation under various operational conditions. Two different methods of salt addition were tested, revealing that the introduction of liquid molten salt led to temperature fluctuations, while the heat pipe start-up process was influenced by the melting of molten salt during cold start-up. During steady power operation, the system exhibited stability, with natural convection of molten salt in the annular gap enhancing heat transfer. The primary factor affecting thermoelectric conversion efficiency was identified as the thermal resistance between the condensation section of the heat pipe and the TEG. With increasing heating temperatures, the wall temperatures of each part of heat pipe rose accordingly, resulting in improving heat transfer efficiency and thermoelectric conversion. This investigation is expected to offer valuable insights for the start-up and operation of micro-MSRs.
{"title":"Experimental research on heat transfer in a molten salt-heat pipe-thermoelectric generator system based on micro-MSR","authors":"Xingwei Chen , Zhizhe Xu , Dai Ye , Yang Zou","doi":"10.1016/j.nucengdes.2025.113911","DOIUrl":"10.1016/j.nucengdes.2025.113911","url":null,"abstract":"<div><div>The heat pipe-cooled micro molten salt reactor (micro-MSR) utilizes heat pipes for transferring the fission energy produced in the core to thermoelectric generators (TEG). In order to assess the heat transfer performance, an integrated experimental setup comprising a molten salt – heat pipe – thermoelectric generator was established. Experiments were carried out to assess the system’s performance during start-up and operation under various operational conditions. Two different methods of salt addition were tested, revealing that the introduction of liquid molten salt led to temperature fluctuations, while the heat pipe start-up process was influenced by the melting of molten salt during cold start-up. During steady power operation, the system exhibited stability, with natural convection of molten salt in the annular gap enhancing heat transfer. The primary factor affecting thermoelectric conversion efficiency was identified as the thermal resistance between the condensation section of the heat pipe and the TEG. With increasing heating temperatures, the wall temperatures of each part of heat pipe rose accordingly, resulting in improving heat transfer efficiency and thermoelectric conversion. This investigation is expected to offer valuable insights for the start-up and operation of micro-MSRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"434 ","pages":"Article 113911"},"PeriodicalIF":1.9,"publicationDate":"2025-02-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143421986","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 : 2025-02-13DOI: 10.1016/j.nucengdes.2025.113904
Yutong Chen , Dalin Zhang , Yue Lin , Di Wang , Zhenyu Feng , Wenxi Tian , S.Z. Qiu , G.H. Su
Apart from the inherent advantages including high thermal efficiency, low operating pressure, stable coolant chemical properties, large heat capacity and easy modularization, Lead-based Fast Reactors (LFRs) are capable of transmuting long-live nuclear wastes while breeding fissile nuclides, therefore it is recognized as one of the most promising Gen-IV reactor concepts. In most existing LFR design schemes, the intermedia loop is not utilized, making the reactor more vulnerable to the risk of Steam Generator Tube Rupture (SGTR) accident. As a result, safety analysis of SGTR accident of LFR has become a major concern over the last decades. In this paper, the key phenomena at different stages of SGTR accident of LFR are categorized, the theoretical and experimental research status are reviewed, and the corresponding countermeasures are suggested. Focused on the four typical development stages of SGTR accident of LFR, namely the pressure wave propagating stage, the multiphase mixture expanding stage, the primary coolant (molten lead) and secondary coolant (usually pressurized water) interacting stage and the steam bubble migration stage, the research methods and recent progress are summarized. Besides, investigations on phenomena like rapid depressurization and two-phase critical flow that simultaneously occur in the secondary side are discussed subsequently. For those issues, the latest research activities, existing problems and future outlooks are demonstrated. This paper could provide useful reference for design and safety analysis issues of LFR SGTR accidents.
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