Pub Date : 2024-10-22DOI: 10.1016/j.nucengdes.2024.113659
Abylay Tangirbergen , Nurlan Amangeldi , Shripad T. Revankar , Gani Yergaliuly
Despite nuclear energy being a clean, sustainable source, its safety is a major concern, especially after the Chernobyl and Fukushima accidents. Designing accident-tolerant fuel (ATF) clad materials is a key solution. This review examines the development and behavior of FeCrAl alloys, a promising ATF cladding candidate, under irradiation. FeCrAl alloys show excellent resistance to high-temperature corrosion and oxidation, but irradiation can significantly alter their mechanical properties. This paper consolidates experimental and theoretical studies on irradiation hardening in FeCrAl alloys, highlighting dislocation loops and Cr-rich α’ precipitates as primary hardening contributors. It discusses compositional adjustments, such as adding oxide dispersion strengthening (ODS) materials, and evaluates advanced techniques to mitigate irradiation-induced damage and improve alloy performance. Theoretical frameworks of irradiation hardening and computer simulation methods are overviewed. This review provides a comprehensive understanding of irradiation hardening mechanisms in FeCrAl alloys and suggests future research directions for enhancing nuclear reactor safety and efficiency.
{"title":"A review of irradiation-induced hardening in FeCrAl alloy systems for accident-tolerant fuel cladding","authors":"Abylay Tangirbergen , Nurlan Amangeldi , Shripad T. Revankar , Gani Yergaliuly","doi":"10.1016/j.nucengdes.2024.113659","DOIUrl":"10.1016/j.nucengdes.2024.113659","url":null,"abstract":"<div><div>Despite nuclear energy being a clean, sustainable source, its safety is a major concern, especially after the Chernobyl and Fukushima accidents. Designing accident-tolerant fuel (ATF) clad materials is a key solution. This review examines the development and behavior of FeCrAl alloys, a promising ATF cladding candidate, under irradiation. FeCrAl alloys show excellent resistance to high-temperature corrosion and oxidation, but irradiation can significantly alter their mechanical properties. This paper consolidates experimental and theoretical studies on irradiation hardening in FeCrAl alloys, highlighting dislocation loops and Cr-rich α’ precipitates as primary hardening contributors. It discusses compositional adjustments, such as adding oxide dispersion strengthening (ODS) materials, and evaluates advanced techniques to mitigate irradiation-induced damage and improve alloy performance. Theoretical frameworks of irradiation hardening and computer simulation methods are overviewed. This review provides a comprehensive understanding of irradiation hardening mechanisms in FeCrAl alloys and suggests future research directions for enhancing nuclear reactor safety and efficiency.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113659"},"PeriodicalIF":1.9,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535543","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 : 2024-10-21DOI: 10.1016/j.nucengdes.2024.113645
Abhitab Bachchan, Dhrumil Ganatra, Subhadip Kirtan , K. Devan
The Indian 40 MWt experimental Fast Breeder Test Reactor (FBTR) operating at Kalpakkam has different systems for fuel pin failure detection. It uses a Gaseous Fission Product Detection (GFPD) system to detect the dry rupture phase of fuel pin clad failure, and it also has a delayed neutron detection (DND) system in each primary loop (east and west) for wet rupture phase detection. In 2011, a series of delayed neutron (DN) signal measurements were performed in FBTR to assess the sensitivity and localisation capabilities of the DND system. A special assembly with 19 perforated fuel pins of natural uranium-nickel metal alloy was used as a fission product source (FPS) for this test. In this paper, an integrated analysis has been carried out to simulate the experimental observations by using both neutronics and thermal hydraulics calculations. The Prompt Recoil Model (PRM) and modified Non-Recoil Model (NRM) with isotopic hold-up time are used to estimate the DN precursor release rate from the perforated fuel pin to the coolant sodium. The time-dependent activity is evaluated considering hydraulic dilution and decay of the DN precursors. To get the hydraulic dilution of DN precursors during their transport to the detector, a 3D CFD analysis of FBTR core with entire pool sodium has been performed using the commercial code ANSYS FLUENT. Monte Carlo modelling of the DND system is done for DN signal estimation by considering the spatial distribution of the DN source around the detectors. Results showed that a modified non-recoil DN precursor release model coupled with the neutronics-hydraulics simulation gives better prediction of DN signal in FBTR, and hence, this methodology can be extended for generating the contrast ratio for core locations where measurements are not performed.
{"title":"An integrated analysis of DND experiments in the Indian experimental Fast Breeder reactor using prompt recoil and modified Non-Recoil DN precursor release models","authors":"Abhitab Bachchan, Dhrumil Ganatra, Subhadip Kirtan , K. Devan","doi":"10.1016/j.nucengdes.2024.113645","DOIUrl":"10.1016/j.nucengdes.2024.113645","url":null,"abstract":"<div><div>The Indian 40 MWt experimental Fast Breeder Test Reactor (FBTR) operating at Kalpakkam has different systems for fuel pin failure detection. It uses a Gaseous Fission Product Detection (GFPD) system to detect the dry rupture phase of fuel pin clad failure, and it also has a delayed neutron detection (DND) system in each primary loop (east and west) for wet rupture phase detection. In 2011, a series of delayed neutron (DN) signal measurements were performed in FBTR to assess the sensitivity and localisation capabilities of the DND system. A special assembly with 19 perforated fuel pins of natural uranium-nickel metal alloy was used as a fission product source (FPS) for this test. In this paper, an integrated analysis has been carried out to simulate the experimental observations by using both neutronics and thermal hydraulics calculations. The Prompt Recoil Model (PRM) and modified Non-Recoil Model (NRM) with isotopic hold-up time are used to estimate the DN precursor release rate from the perforated fuel pin to the coolant sodium. The time-dependent activity is evaluated considering hydraulic dilution and decay of the DN precursors. To get the hydraulic dilution of DN precursors during their transport to the detector, a 3D CFD analysis of FBTR core with entire pool sodium has been performed using the commercial code ANSYS FLUENT. Monte Carlo modelling of the DND system is done for DN signal estimation by considering the spatial distribution of the DN source around the detectors. Results showed that a modified non-recoil DN precursor release model coupled with the neutronics-hydraulics simulation gives better prediction of DN signal in FBTR, and hence, this methodology can be extended for generating the contrast ratio for core locations where measurements are not performed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113645"},"PeriodicalIF":1.9,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535514","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 : 2024-10-21DOI: 10.1016/j.nucengdes.2024.113636
Sinem Uzun, Eyyüp Yildiz, Hatice Arslantaş
This study emphasizes how important accurate prediction of channel temperatures in nuclear reactors is for safety and operational efficiency. While traditional methods require long and complex processes such as kernel modeling and mathematical simulations, artificial neural networks (ANN) provide more efficient predictions by accelerating this process. The superior ability of ANNs to process large data sets is intended to demonstrate that this study will provide a valuable alternative compared to conventional methods and increase the accuracy of reactor temperature predictions. In this study, the training performances of Artificial Neural Network (ANN) developed to determine the nuclear reactor channel temperature with different hyperparameter combinations were analysed. It was conducted several experimental studies to assess the influence of hyperparameters on our model for nuclear reactor parameter data prediction. The training and validation results indicates that learning rate, hidden layer sizes and number have critical effects for the more precisive prediction. It was observed that models with a learning rate of 0.05 and 0.5 achieved successful learning with less fluctuation in training and validation errors. When looking at hidden layer sizes, networks with 32 and 64 neurons performed better than networks with 16 neurons. For the test phase our model can successfully predict data with slight error margin. As a result, we demonstrated that neural networks are a powerful tool in nuclear reactor channel temperature prediction through our proposed model.
{"title":"Optimizing neural network models for predicting nuclear reactor channel temperature: A study on hyperparameter tuning and performance analysis","authors":"Sinem Uzun, Eyyüp Yildiz, Hatice Arslantaş","doi":"10.1016/j.nucengdes.2024.113636","DOIUrl":"10.1016/j.nucengdes.2024.113636","url":null,"abstract":"<div><div>This study emphasizes how important accurate prediction of channel temperatures in nuclear reactors is for safety and operational efficiency. While traditional methods require long and complex processes such as kernel modeling and mathematical simulations, artificial neural networks (ANN) provide more efficient predictions by accelerating this process. The superior ability of ANNs to process large data sets is intended to demonstrate that this study will provide a valuable alternative compared to conventional methods and increase the accuracy of reactor temperature predictions. In this study, the training performances of Artificial Neural Network (ANN) developed to determine the nuclear reactor channel temperature with different hyperparameter combinations were analysed. It was conducted several experimental studies to assess the influence of hyperparameters on our model for nuclear reactor parameter data prediction. The training and validation results indicates that learning rate, hidden layer sizes and number have critical effects for the more precisive prediction. It was observed that models with a learning rate of 0.05 and 0.5 achieved successful learning with less fluctuation in training and validation errors. When looking at hidden layer sizes, networks with 32 and 64 neurons performed better than networks with 16 neurons. For the test phase our model can successfully predict data with slight error margin. As a result, we demonstrated that neural networks are a powerful tool in nuclear reactor channel temperature prediction through our proposed model.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113636"},"PeriodicalIF":1.9,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535658","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 : 2024-10-21DOI: 10.1016/j.nucengdes.2024.113639
Thomas Guilbaud , Eymeric Simonnot , Alessandro Scolaro , Carlo Fiorina
Nuclear Thermal Propulsion systems are one of the technologies to unlock interplanetary travel in the solar system and there is currently an increasing demand for accurately simulating such systems. The present paper investigates the KIWI-B-4E reactor, the first prototype of the NERVA program that achieved a restart at full power in 1964. To simulate this reactor, we propose the use of the OpenMC Monte Carlo code and the GeN-Foam multi-physics code to model full-core neutronics and thermal-hydraulics. The results mostly show a good agreement with the available experimental data, although uncertainties on key design parameters do not allow for a detailed neutronics validation.
核热推进系统是开启太阳系星际旅行的技术之一，目前对精确模拟此类系统的需求与日俱增。本文研究了 KIWI-B-4E 反应堆，它是 NERVA 计划的第一个原型，于 1964 年实现了全功率重启。为了模拟该反应堆，我们建议使用 OpenMC Monte Carlo 代码和 GeN-Foam 多物理场代码来模拟全堆芯中子电子学和热水力学。尽管关键设计参数的不确定性导致无法进行详细的中子验证，但大部分结果显示与现有实验数据非常吻合。
{"title":"Full core study of the KIWI-B-4E Nuclear Thermal Propulsion system using OpenMC and GeN-Foam","authors":"Thomas Guilbaud , Eymeric Simonnot , Alessandro Scolaro , Carlo Fiorina","doi":"10.1016/j.nucengdes.2024.113639","DOIUrl":"10.1016/j.nucengdes.2024.113639","url":null,"abstract":"<div><div>Nuclear Thermal Propulsion systems are one of the technologies to unlock interplanetary travel in the solar system and there is currently an increasing demand for accurately simulating such systems. The present paper investigates the KIWI-B-4E reactor, the first prototype of the NERVA program that achieved a restart at full power in 1964. To simulate this reactor, we propose the use of the OpenMC Monte Carlo code and the GeN-Foam multi-physics code to model full-core neutronics and thermal-hydraulics. The results mostly show a good agreement with the available experimental data, although uncertainties on key design parameters do not allow for a detailed neutronics validation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113639"},"PeriodicalIF":1.9,"publicationDate":"2024-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535659","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 : 2024-10-20DOI: 10.1016/j.nucengdes.2024.113638
Kenichi Hakozaki
International Atomic Energy Agency (IAEA) regulations for the safe transport of radioactive materials require a 9-m drop test for the Type B packages, such as large spent fuel (SF) transport casks. In vertical drop tests, where the SF cask is in a lid-down orientation, it has been observed that the pullout force of the lid bolts delays the impact from the grounding of the specimen. This phenomenon is known as delayed impact and is of significant concern for the safe transport of radioactive materials. Previously, the delayed impact has been considered to occur because of the delayed collision of the contents in the cask with its lid when there is a gap between the contents and lid. In such case, since the lid of the SF cask is usually fixed at the end of the cask body, the contents push the lid plate to increase the pullout force, leading to the delayed impact. Herein, a new noise reduction method, which is the acceleration integrated velocity approximation (AIVA) method, is proposed to analyze the acceleration data from a 9-m drop test. Based on a reevaluation of the existing acceleration data using the proposed noise reduction method, it is concluded that the delayed impact could occur even when there is no gap between the contents and lid. To evaluate practical SF casks, the mechanical characteristics of the contents are estimated based on the published results of a full-scale drop test, and the delayed impact in a no-gap case is estimated based on the obtained characteristics. In a previous safety analysis of a wet-type SF cask, the delayed impact was not considered to be important, since the contents might be restricted by the water present inside the cask. However, based on this study, it is concluded that the delayed impact caused by the content reaction should be considered even in such no gap cases.
国际原子能机构（IAEA）关于放射性物质安全运输的规定要求对 B 型包件（如大型乏燃料（SF）运输桶）进行 9 米跌落试验。在垂直跌落试验中，乏燃料运输桶的盖子是朝下的，据观察，盖子螺栓的拉力会延迟试样接地产生的冲击力。这种现象被称为延迟冲击，是放射性物质安全运输的重大问题。以前，人们认为发生延迟撞击的原因是，当桶内物品与桶盖之间存在间隙时，桶内物品与桶盖会发生延迟碰撞。在这种情况下，由于 SF 掩埋桶的桶盖通常固定在桶体的末端，内装物会推动桶盖板以增加拉拔力，从而导致延迟撞击。本文提出了一种新的降噪方法，即加速度综合速度近似法（AIVA），用于分析 9 米跌落试验的加速度数据。在使用所提出的降噪方法重新评估现有加速度数据的基础上，得出的结论是，即使内装物和盖子之间没有间隙，也可能发生延迟冲击。为了评估实用的 SF 桶，根据已公布的全尺寸跌落试验结果估算了内装物的机械特性，并根据获得的特性估算了无间隙情况下的延迟撞击。在以前对湿式 SF 罐进行的安全分析中，延迟冲击被认为并不重要，因为罐内物品可能会受到罐内水的限制。然而，根据这项研究得出的结论是，即使在这种无间隙情况下，也应考虑内装物反应引起的延迟影响。
{"title":"Delay impact mechanism of Spent Fuel Cask without assuming content dropping inside packaging","authors":"Kenichi Hakozaki","doi":"10.1016/j.nucengdes.2024.113638","DOIUrl":"10.1016/j.nucengdes.2024.113638","url":null,"abstract":"<div><div>International Atomic Energy Agency (IAEA) regulations for the safe transport of radioactive materials require a 9-m drop test for the Type B packages, such as large spent fuel (SF) transport casks. In vertical drop tests, where the SF cask is in a lid-down orientation, it has been observed that the pullout force of the lid bolts delays the impact from the grounding of the specimen. This phenomenon is known as delayed impact and is of significant concern for the safe transport of radioactive materials. Previously, the delayed impact has been considered to occur because of the delayed collision of the contents in the cask with its lid when there is a gap between the contents and lid. In such case, since the lid of the SF cask is usually fixed at the end of the cask body, the contents push the lid plate to increase the pullout force, leading to the delayed impact. Herein, a new noise reduction method, which is the acceleration integrated velocity approximation (AIVA) method, is proposed to analyze the acceleration data from a 9-m drop test. Based on a reevaluation of the existing acceleration data using the proposed noise reduction method, it is concluded that the delayed impact could occur even when there is no gap between the contents and lid. To evaluate practical SF casks, the mechanical characteristics of the contents are estimated based on the published results of a full-scale drop test, and the delayed impact in a no-gap case is estimated based on the obtained characteristics. In a previous safety analysis of a wet-type SF cask, the delayed impact was not considered to be important, since the contents might be restricted by the water present inside the cask. However, based on this study, it is concluded that the delayed impact caused by the content reaction should be considered even in such no gap cases.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113638"},"PeriodicalIF":1.9,"publicationDate":"2024-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535661","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 : 2024-10-19DOI: 10.1016/j.nucengdes.2024.113633
Hyunwoo Yook, Sunghoon Joung, Chansoo Lee, Youho Lee
This paper introduces the Integral Loss Of Coolant (LOCA) facility (i-LOCA) established at Seoul National University. The facility was designed to investigate the integral fuel behavior of Light Water Reactors during LOCA, encompassing aspects such as cladding oxidation, ballooning and burst, reflood quenching, secondary hydriding, and fuel pellet dispersal. Integral LOCA experiments were carried out using three types of surrogate ZrO2 pellets, representing various segment burnups: cylindrical pellets with no fuel fragmentation (<55 GWd/MTU), mixed fragments of different sizes simulating ∼68 GWd/MTU (D = 0.3, 0.5, 1.0, 2.0, 3.0, and 5.0 mm with the same mass fraction), and small single powdered fragments simulating ultra-high burnup fuel (D = 0.5 mm, ∼94 GWd/MTU). Zr-Nb-Sn, Zr-1.1Nb, and Cr-coated (15 μm, Arc Ion Plating) Zr-1.1Nb ATF cladding were employed, with rod internal pressures ranging from 1 MPa to 7 MPa. The burst size and hoop strain exhibited significant variations depending on the type of surrogate pellets used, with larger burst sizes and hoop strains observed for smaller average diameters of surrogate pellets due to the effect of azimuthal and axial temperature distribution. Fuel dispersal was influenced by rod internal pressure, burst size, and the size of pellet fragments. Only pellet fragments smaller than the burst hole width underwent dispersal upon fuel burst, while larger fragments blocked the dispersal of smaller fragments. The rapidly escalating average dispersal fraction of single powder compared to mixed powder indicated a threshold burnup for fuel dispersal between 69–94 GWd/MTU. Cladding inner oxidation length was influenced by burst hole size and remaining fuel pellets. The results of inner oxidation confirmed the validity of the U.S. NRC’s assumption regarding the length of inner wall oxidation. The tested 15 μm Cr-coated cladding tubes, produced using the arc ion plating method, exhibited no significant differences in burst geometry, fuel dispersal, inner oxidation, and secondary hydriding when compared to the uncoated reference cladding.
{"title":"Integral LOCA experiments to study FFRD behavior of high burnup nuclear fuels","authors":"Hyunwoo Yook, Sunghoon Joung, Chansoo Lee, Youho Lee","doi":"10.1016/j.nucengdes.2024.113633","DOIUrl":"10.1016/j.nucengdes.2024.113633","url":null,"abstract":"<div><div>This paper introduces the Integral Loss Of Coolant (LOCA) facility (i-LOCA) established at Seoul National University. The facility was designed to investigate the integral fuel behavior of Light Water Reactors during LOCA, encompassing aspects such as cladding oxidation, ballooning and burst, reflood quenching, secondary hydriding, and fuel pellet dispersal. Integral LOCA experiments were carried out using three types of surrogate ZrO<sub>2</sub> pellets, representing various segment burnups: cylindrical pellets with no fuel fragmentation (<55 GWd/MTU), mixed fragments of different sizes simulating ∼68 GWd/MTU (D = 0.3, 0.5, 1.0, 2.0, 3.0, and 5.0 mm with the same mass fraction), and small single powdered fragments simulating ultra-high burnup fuel (D = 0.5 mm, ∼94 GWd/MTU). Zr-Nb-Sn, Zr-1.1Nb, and Cr-coated (15 μm, Arc Ion Plating) Zr-1.1Nb ATF cladding were employed, with rod internal pressures ranging from 1 MPa to 7 MPa. The burst size and hoop strain exhibited significant variations depending on the type of surrogate pellets used, with larger burst sizes and hoop strains observed for smaller average diameters of surrogate pellets due to the effect of azimuthal and axial temperature distribution. Fuel dispersal was influenced by rod internal pressure, burst size, and the size of pellet fragments. Only pellet fragments smaller than the burst hole width underwent dispersal upon fuel burst, while larger fragments blocked the dispersal of smaller fragments. The rapidly escalating average dispersal fraction of single powder compared to mixed powder indicated a threshold burnup for fuel dispersal between 69–94 GWd/MTU. Cladding inner oxidation length was influenced by burst hole size and remaining fuel pellets. The results of inner oxidation confirmed the validity of the U.S. NRC’s assumption regarding the length of inner wall oxidation. The tested 15 μm Cr-coated cladding tubes, produced using the arc ion plating method, exhibited no significant differences in burst geometry, fuel dispersal, inner oxidation, and secondary hydriding when compared to the uncoated reference cladding.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113633"},"PeriodicalIF":1.9,"publicationDate":"2024-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142535660","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 : 2024-10-15DOI: 10.1016/j.nucengdes.2024.113640
Luca Berti, Donato Aquaro, Rosa Lo Frano
In case of an in-vessel Loss Of Coolant Accident (LOCA), flash steam can be released in the Vacuum Vessel (VV) of the International Thermonuclear Experimental Reactor (ITER) causing its pressurization. To avoid this, the safety system named Vacuum Vessel Pressure Suppression System (VVPSS) will intervene sending the steam to four Vapour Suppression Tanks (VSTs) through a multi-hole sparger and condenses via Direct Contact Condensation (DCC).
To support the design of the multi-hole sparger, which is a key safety component of VVPSS, at the University of Pisa two testing facilities were designed and built in order to study and qualify the VVPSS, named Small Scale Test Facility (SSTF) and Large Scale Test Facility (LSTF).
During the experimental tests performed using LSTF with a VVPSS multi-hole full scale sparger, under certain conditions, the coalescence of the steam jet plumes resulted in the formation and collapse of large, isolated steam bubbles which produced high pressure loads at low frequency on the structure and flow reversal of the pool water inside the sparger.
To limit these large pressure loads, a methodology is needed to prevent the coalescence of the steam jet plumes.
With this aim, an image analysis of 15 experimental tests performed using SSTF was performed to develop and validate a correlation of the ratio between the maximum radius of the steam jet plumes and the hole diameter. Subsequently, two limiting radii for multi-hole spargers (named r1 and r2) were determined which allow avoiding the partial and the transitional complete coalescence of the steam jet plumes when compared to the maximum radius. The proposed methodology is new and quite innovative, and it was applied and validated by using the several videos recorded during the transient test performed using sparger B, consisting of DN450 pipe with 1000 holes.
The correlation estimates that partial coalescence and transitional to complete coalescence regions are avoided when the water subcooling temperature ranges between 37–45 °C and 25–31 °C, respectively, as observed in the recordings of the cameras. Results allow to identify the sparger design dimensions preventing the steam jet plumes coalescence, and avoiding the onset of excessive dynamic loads.
{"title":"Design methodology of multi-hole spargers to prevent steam coalescence at sub-atmospheric pressure","authors":"Luca Berti, Donato Aquaro, Rosa Lo Frano","doi":"10.1016/j.nucengdes.2024.113640","DOIUrl":"10.1016/j.nucengdes.2024.113640","url":null,"abstract":"<div><div>In case of an in-vessel Loss Of Coolant Accident (LOCA), flash steam can be released in the Vacuum Vessel (VV) of the International Thermonuclear Experimental Reactor (ITER) causing its pressurization. To avoid this, the safety system named Vacuum Vessel Pressure Suppression System (VVPSS) will intervene sending the steam to four Vapour Suppression Tanks (VSTs) through a multi-hole sparger and condenses via Direct Contact Condensation (DCC).</div><div>To support the design of the multi-hole sparger, which is a key safety component of VVPSS, at the University of Pisa two testing facilities were designed and built in order to study and qualify the VVPSS, named Small Scale Test Facility (SSTF) and Large Scale Test Facility (LSTF).</div><div>During the experimental tests performed using LSTF with a VVPSS multi-hole full scale sparger, under certain conditions, the coalescence of the steam jet plumes resulted in the formation and collapse of large, isolated steam bubbles which produced high pressure loads at low frequency on the structure and flow reversal of the pool water inside the sparger.</div><div>To limit these large pressure loads, a methodology is needed to prevent the coalescence of the steam jet plumes.</div><div>With this aim, an image analysis of 15 experimental tests performed using SSTF was performed to develop and validate a correlation of the ratio between the maximum radius of the steam jet plumes and the hole diameter. Subsequently, two limiting radii for multi-hole spargers (named r<sub>1</sub> and r<sub>2</sub>) were determined which allow avoiding the partial and the transitional complete coalescence of the steam jet plumes when compared to the maximum radius. The proposed methodology is new and quite innovative, and it was applied and validated by using the several videos recorded during the transient test performed using sparger B, consisting of DN450 pipe with 1000 holes.</div><div>The correlation estimates that partial coalescence and transitional to complete coalescence regions are avoided when the water subcooling temperature ranges between 37–45 °C and 25–31 °C, respectively, as observed in the recordings of the cameras. Results allow to identify the sparger design dimensions preventing the steam jet plumes coalescence, and avoiding the onset of excessive dynamic loads.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113640"},"PeriodicalIF":1.9,"publicationDate":"2024-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142437838","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 : 2024-10-15DOI: 10.1016/j.nucengdes.2024.113620
Chenrui Mao, Baihui Jiang, Yu Ji, Jun Sun, Lei Shi
The Nuclear Thermal Propulsion (NTP) boasts advantages such as high specific impulse, substantial thrust, and extended operating time, giving it a clear edge in deep space exploration and orbital maneuvers. To fully harness the potential of NTP, transient analysis is crucial to ensure reliability, safety, and performance under various operational conditions. In this paper, a transient version of NTP analysis code PANES-Tran (Program for Analyzing Nuclear Engine Systems − Transient) was developed for the 110kN expander cycle particle bed reactor (PBR) nuclear thermal propulsion (NTP) system. The code is based on the one-dimensional thermal-hydraulics (TH) framework and its fixed-point iteration expressions, coupled with the point reactor kinetics (PK) model. Under the framework, a turbopump model incorporating characteristic curves was constructed, and a PBR fuel element model involving porous media and fuel particle heat transfer procedure was also established. The basic models and methods were preliminary verified using AMESim for fluid flow and heat transfer and, RELAP5 for PK/TH coupling scheme. Moreover, the integrated effect of PANES-Tran was also verified by the design parameters of the 110 kN PBR-NTP system. Subsequently, a transient process triggered by +0.2$ step reactivity introduction under rated conditions was studied, which indicated that the NTP system could stably transition to a new steady state with a thrust of 125 kN. This study could provide a powerful tool for subsequent research on transient characteristics and operation strategy for NTP systems.
{"title":"Development and preliminary verification of a transient analysis code PANES-Tran for Nuclear thermal propulsion","authors":"Chenrui Mao, Baihui Jiang, Yu Ji, Jun Sun, Lei Shi","doi":"10.1016/j.nucengdes.2024.113620","DOIUrl":"10.1016/j.nucengdes.2024.113620","url":null,"abstract":"<div><div>The Nuclear Thermal Propulsion (NTP) boasts advantages such as high specific impulse, substantial thrust, and extended operating time, giving it a clear edge in deep space exploration and orbital maneuvers. To fully harness the potential of NTP, transient analysis is crucial to ensure reliability, safety, and performance under various operational conditions. In this paper, a transient version of NTP analysis code PANES-Tran (Program for Analyzing Nuclear Engine Systems − Transient) was developed for the 110kN expander cycle particle bed reactor (PBR) nuclear thermal propulsion (NTP) system. The code is based on the one-dimensional thermal-hydraulics (TH) framework and its fixed-point iteration expressions, coupled with the point reactor kinetics (PK) model. Under the framework, a turbopump model incorporating characteristic curves was constructed, and a PBR fuel element model involving porous media and fuel particle heat transfer procedure was also established. The basic models and methods were preliminary verified using AMESim for fluid flow and heat transfer and, RELAP5 for PK/TH coupling scheme. Moreover, the integrated effect of PANES-Tran was also verified by the design parameters of the 110 kN PBR-NTP system. Subsequently, a transient process triggered by +0.2$ step reactivity introduction under rated conditions was studied, which indicated that the NTP system could stably transition to a new steady state with a thrust of 125 kN. This study could provide a powerful tool for subsequent research on transient characteristics and operation strategy for NTP systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113620"},"PeriodicalIF":1.9,"publicationDate":"2024-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142437839","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 : 2024-10-15DOI: 10.1016/j.nucengdes.2024.113647
Tianbao Lan , Tianyou Feng , Feng Sheng , Wei Tan
The structural integrity of core graphite under high-temperature irradiation conditions is crucial for the safe operation of the reactor. This paper presents a simulation of the hexagonal core graphite structure, developed using the UMAT program. Design factors, including irradiation, high temperature, dimensional strain, and creep strain, are analyzed separately through the control variable method. The results indicate a positive correlation between the magnitude of the irradiation field gradient and the resulting stress effects. Stress concentration within the temperature field is observed to occur near the inner side of the hexagonal prism. Among the four types of graphite examined, PCIB demonstrates the least stress and deformation, making it more suitable for specific applications. The selection of graphite should consider the particular service period requirements. Choosing a graphite material that exhibits minimal shrinkage and a high turnaround dose is advisable. The primary creep parameter is negligible when compared to the secondary creep parameter; selecting graphite with a larger secondary creep parameter enhances reactor safety. The findings of this study provide a solid foundation for the design of a graphite core and offer recommendations for graphite candidates in the development of microreactors in China.
高温辐照条件下堆芯石墨的结构完整性对反应堆的安全运行至关重要。本文介绍了利用 UMAT 程序开发的六边形堆芯石墨结构模拟。通过控制变量法分别分析了辐照、高温、尺寸应变和蠕变应变等设计因素。结果表明，辐照场梯度的大小与所产生的应力效应呈正相关。据观察，温度场中的应力集中发生在六棱柱内侧附近。在所研究的四种石墨中，PCIB 的应力和变形最小，因此更适合特定应用。选择石墨时应考虑特定的使用期限要求。最好选择收缩率最小、周转率高的石墨材料。与二次蠕变参数相比，一次蠕变参数可忽略不计；选择二次蠕变参数较大的石墨可提高反应堆的安全性。本研究的结果为石墨堆芯的设计奠定了坚实的基础，并为中国开发微反应器的石墨候选材料提供了建议。
{"title":"Internal stress analysis of irradiated graphite cores in a gas-cooled microreactor","authors":"Tianbao Lan , Tianyou Feng , Feng Sheng , Wei Tan","doi":"10.1016/j.nucengdes.2024.113647","DOIUrl":"10.1016/j.nucengdes.2024.113647","url":null,"abstract":"<div><div>The structural integrity of core graphite under high-temperature irradiation conditions is crucial for the safe operation of the reactor. This paper presents a simulation of the hexagonal core graphite structure, developed using the UMAT program. Design factors, including irradiation, high temperature, dimensional strain, and creep strain, are analyzed separately through the control variable method. The results indicate a positive correlation between the magnitude of the irradiation field gradient and the resulting stress effects. Stress concentration within the temperature field is observed to occur near the inner side of the hexagonal prism. Among the four types of graphite examined, PCIB demonstrates the least stress and deformation, making it more suitable for specific applications. The selection of graphite should consider the particular service period requirements. Choosing a graphite material that exhibits minimal shrinkage and a high turnaround dose is advisable. The primary creep parameter is negligible when compared to the secondary creep parameter; selecting graphite with a larger secondary creep parameter enhances reactor safety. The findings of this study provide a solid foundation for the design of a graphite core and offer recommendations for graphite candidates in the development of microreactors in China.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113647"},"PeriodicalIF":1.9,"publicationDate":"2024-10-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142442293","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 : 2024-10-14DOI: 10.1016/j.nucengdes.2024.113635
Muping Li, Aodi Sun, Peiwei Sun, Xinyu Wei
Small lead–bismuth fast reactor(SLBFR) with pool cooling offers inherent safety features, making it suitable for diverse applications such as mobile nuclear power plants and remote power supplies However, conventional control methods often struggle to meet the demands of frequent load adjustments. To ensure the safe operation of SLBFR in the turbine-leading mode, a new control method for the once-through steam generator (OTSG) is imperative. Moreover, enhancing load-following capabilities is essential to meet operational requirements. The model of SLBFR is established in MATLAB/Simulink software to study the characteristics of OTSG operating in the turbine-leading mode. The control system of steam pressure is built by active disturbance rejection control. The load-following characteristic of core power is improved by load feedforward control modified by energy balance. Simulation results demonstrate that the adopted control method enhances the load-following capability of the OTSG. This paper serves as a valuable reference for designing the OTSG control system of SLBFR, contributing to its safe and efficient operation in various settings.
{"title":"Control method of once-through stream generator based on active disturbance rejection control","authors":"Muping Li, Aodi Sun, Peiwei Sun, Xinyu Wei","doi":"10.1016/j.nucengdes.2024.113635","DOIUrl":"10.1016/j.nucengdes.2024.113635","url":null,"abstract":"<div><div>Small lead–bismuth fast reactor(SLBFR) with pool cooling offers inherent safety features, making it suitable for diverse applications such as mobile nuclear power plants and remote power supplies However, conventional control methods often struggle to meet the demands of frequent load adjustments. To ensure the safe operation of SLBFR in the turbine-leading mode, a new control method for the once-through steam generator (OTSG) is imperative. Moreover, enhancing load-following capabilities is essential to meet operational requirements. The model of SLBFR is established in MATLAB/Simulink software to study the characteristics of OTSG operating in the turbine-leading mode. The control system of steam pressure is built by active disturbance rejection control. The load-following characteristic of core power is improved by load feedforward control modified by energy balance. Simulation results demonstrate that the adopted control method enhances the load-following capability of the OTSG. This paper serves as a valuable reference for designing the OTSG control system of SLBFR, contributing to its safe and efficient operation in various settings.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"429 ","pages":"Article 113635"},"PeriodicalIF":1.9,"publicationDate":"2024-10-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142434310","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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