F. Hernández, P. Arena, L. Boccaccini, I. Cristescu, A. Del Nevo, Pierre Sardain, G. Spagnuolo, M. Utili, A. Venturini, Guangming Zhou
The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021.
{"title":"Advancements in Designing the DEMO Driver Blanket System at the EU DEMO Pre-Conceptual Design Phase: Overview, Challenges and Opportunities","authors":"F. Hernández, P. Arena, L. Boccaccini, I. Cristescu, A. Del Nevo, Pierre Sardain, G. Spagnuolo, M. Utili, A. Venturini, Guangming Zhou","doi":"10.3390/jne4030037","DOIUrl":"https://doi.org/10.3390/jne4030037","url":null,"abstract":"The EU conducted the pre-conceptual design (PCD) phase of the demonstration reactor (DEMO) during 2014–2020 under the framework of the EUROfusion consortium. The current strategy of DEMO design is to bridge the breeding blanket (BB) technology gaps between ITER and a commercial fusion power plant (FPP) by playing the role of a “Component Test Facility” for the BB. Within this strategy, a so-called driver blanket, with nearly full in-vessel surface coverage, will aim at achieving high-level stakeholder requirements of tritium self-sufficiency and power extraction for net electricity production with rather conventional technology and/or operational parameters, while an advanced blanket (or several of them) will aim at demonstrating, with limited coverage, features that are deemed necessary for a commercial FPP. Currently, two driver blanket candidates are being investigated for the EU DEMO, namely the water-cooled lithium lead and the helium-cooled pebble bed breeding blanket concepts. The PCD phase has been characterized not only by the detailed design of the BB systems themselves, but also by their holistic integration in DEMO, prioritizing near-term solutions, in accordance with the idea of a driver blanket. This paper summarizes the status for both BB driver blanket candidates at the end of the PCD phase, including their corresponding tritium extraction and removal (TER) systems, underlining the main achievements and lessons learned, exposing outstanding key system design and R&D challenges and presenting identified opportunities to address those risks during the conceptual design (CD) phase that started in 2021.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-08-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83332011","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect, but up to now, no experimental program has confirmed this effect. The French Plutonium Temperature Effect Experimental Program (or PU+ in short) aims to effectively show that such a positive temperature effect exists for diluted plutonium solutions. The PU+ experiments were conducted in the “Apparatus B” facility at the CEA VALDUC research center in France. It involved several sub-critical approach-type experiments using plutonium nitrate solutions with concentrations of 14.3, 15, and 20 g/L at temperatures ranging from 20 to 40 °C. Fourteen (five at 20 g/L, four at 15 g/L, and five at 14.3 g/L) phase I experiments (consisting of independent sub-critical approaches) were performed between 2006 and 2007. The impact of the uncertainties on solution acidity and plutonium concentration made it difficult to demonstrate the positive temperature effect, requiring an additional phase II experiment (with a unique plutonium solution) from 22 to 28 °C that was performed in July 2007. This phase II experiment has shown the existence of a positive temperature effect of ~+5.17 pcm/°C (from 22 to 28 °C for a plutonium concentration of 14.3 g/L). It has recently been possible to confirm the results of this program with MORET 5 calculations by generating thermal scattering data S(α,β) at the correct experimental temperatures. This paper finally presents a fully documented experimental program highlighting the Plutonium Temperature Effect theoretically described in the literature. Its high level of precision and its “one-step” approach to criticality allowed it to show a significant positive temperature effect for a rather small variation of temperature (+6 °C). The order of magnitude of the effect was confirmed with Monte Carlo calculations using thermal scattering data for hydrogen in the solution produced by IRSN for the purpose of the comparison.
{"title":"The Plutonium Temperature Effect Program","authors":"N. Leclaire, Vaibhav Jaiswal","doi":"10.3390/jne4030035","DOIUrl":"https://doi.org/10.3390/jne4030035","url":null,"abstract":"Various theoretical studies have shown that highly diluted plutonium solutions could have a positive temperature effect, but up to now, no experimental program has confirmed this effect. The French Plutonium Temperature Effect Experimental Program (or PU+ in short) aims to effectively show that such a positive temperature effect exists for diluted plutonium solutions. The PU+ experiments were conducted in the “Apparatus B” facility at the CEA VALDUC research center in France. It involved several sub-critical approach-type experiments using plutonium nitrate solutions with concentrations of 14.3, 15, and 20 g/L at temperatures ranging from 20 to 40 °C. Fourteen (five at 20 g/L, four at 15 g/L, and five at 14.3 g/L) phase I experiments (consisting of independent sub-critical approaches) were performed between 2006 and 2007. The impact of the uncertainties on solution acidity and plutonium concentration made it difficult to demonstrate the positive temperature effect, requiring an additional phase II experiment (with a unique plutonium solution) from 22 to 28 °C that was performed in July 2007. This phase II experiment has shown the existence of a positive temperature effect of ~+5.17 pcm/°C (from 22 to 28 °C for a plutonium concentration of 14.3 g/L). It has recently been possible to confirm the results of this program with MORET 5 calculations by generating thermal scattering data S(α,β) at the correct experimental temperatures. This paper finally presents a fully documented experimental program highlighting the Plutonium Temperature Effect theoretically described in the literature. Its high level of precision and its “one-step” approach to criticality allowed it to show a significant positive temperature effect for a rather small variation of temperature (+6 °C). The order of magnitude of the effect was confirmed with Monte Carlo calculations using thermal scattering data for hydrogen in the solution produced by IRSN for the purpose of the comparison.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-08-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86741510","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The present study investigated the release of tritium from beryllium irradiated at 323 K to a neutron fluence of 4.67 × 1026 m−2 (E > 1 MeV), corresponding up to 22,000 appm helium and 2000 appm tritium productions. The TPD tests revealed a single tritium release peak during thermal desorption tests, irrespective of the heating mode employed. The tritium release peaks occurred at temperatures ranging from 1031–1136 K, depending on the heating mode, with a desorption energy of 1.6 eV. Additionally, the effective tritium diffusion coefficient was found to vary from 1.2 × 10−12 m2/s at 873 K to 1.8 × 10−10 m2/s at 1073 K. The evolution of beryllium microstructure was found to be dependent on the annealing temperature. No discernible differences were observed between the as-received state and after annealing at 473–773 K for 5 h, with a corresponding porosity range of 1–2%. The annealing at temperatures of 873–1373 K for 5 h resulted in the formation of large bubbles, with porosity increasing sharply above 873 K and reaching 30–60%.
{"title":"Tritium Desorption Behavior and Microstructure Evolution of Beryllium Irradiated at Low Temperature Up to High Neutron Dose in BR2 Reactor","authors":"V. Chakin, R. Rolli, R. Gaisin, W. Van Renterghem","doi":"10.3390/jne4030036","DOIUrl":"https://doi.org/10.3390/jne4030036","url":null,"abstract":"The present study investigated the release of tritium from beryllium irradiated at 323 K to a neutron fluence of 4.67 × 1026 m−2 (E > 1 MeV), corresponding up to 22,000 appm helium and 2000 appm tritium productions. The TPD tests revealed a single tritium release peak during thermal desorption tests, irrespective of the heating mode employed. The tritium release peaks occurred at temperatures ranging from 1031–1136 K, depending on the heating mode, with a desorption energy of 1.6 eV. Additionally, the effective tritium diffusion coefficient was found to vary from 1.2 × 10−12 m2/s at 873 K to 1.8 × 10−10 m2/s at 1073 K. The evolution of beryllium microstructure was found to be dependent on the annealing temperature. No discernible differences were observed between the as-received state and after annealing at 473–773 K for 5 h, with a corresponding porosity range of 1–2%. The annealing at temperatures of 873–1373 K for 5 h resulted in the formation of large bubbles, with porosity increasing sharply above 873 K and reaching 30–60%.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-08-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79642234","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The heat transfer during pool boiling incorporates a higher rate of heat dissipation capability at low-temperature differences. This technique is widely used in the nuclear industry for thermal management. In this study, the effect of tube diameter and length on critical heat flux (CHF) at atmospheric conditions in saturated water during pool boiling is analyzed. The tubes of SS 304 are kept in the vertical orientation. The diameter of the tubes ranges from 1.2 mm to 9 mm. The tube lengths varying from 50mm to 1000mm. It has been noted that tubes of smaller diameter show a greater magnitude of CHF for the given length. For a given diameter, a longer tube is found to have lower CHF than the ones having lesser length. The variation in the CHF magnitude is negligible for tubes with a diameter of more than 2.5 mm beyond a length of 200 mm. The location of occurrence of CHF is near the bottom end of the vertical tube. The study illustrates the behavior of bubble nucleation for various tube dimensions and heat fluxes. The inception and detachment of bubbles for different tubes are analyzed. The pool boiling regime is categorized and studied basing the behaviour of the incepted and departed bubble. A mathematical relation that empirically accounts for the effect of tube dimensions i.e. length and diameter on pool boiling CHF is proposed. The experimental CHF data obtained during pool boiling are tabulated towards contributing to the CHF databank.
池沸腾过程中的传热在低温差下具有较高的散热能力。该技术广泛应用于核工业的热管理。本文分析了常压条件下饱和水池沸腾过程中管径和管长对临界热流密度的影响。不锈钢管保持在垂直方向。管的直径范围为1.2 mm ~ 9mm。管长从50mm到1000mm不等。已经注意到，对于给定的长度，直径较小的管显示出更大的CHF幅度。对于给定的直径，较长的管子比长度较小的管子具有更低的CHF。对于直径大于2.5 mm且长度超过200 mm的管，CHF大小的变化可以忽略不计。CHF发生的位置在垂直管的下端附近。研究说明了不同管径和不同热流密度下的气泡成核行为。分析了不同管道中气泡的产生和分离。根据入泡和离泡的行为对池沸腾状态进行了分类和研究。本文提出了一种计算管道尺寸(即长度和直径)对池沸腾CHF影响的数学关系。池沸腾过程中获得的实验CHF数据被制成表格，以供CHF数据库使用。
{"title":"Experimental Analysis Of Bubble Behavior And Critical Heat Flux During Pool Boiling On Vertical Circular Tubes","authors":"Bikash Pattanayak, Hardik B. Kothadia","doi":"10.1115/1.4063041","DOIUrl":"https://doi.org/10.1115/1.4063041","url":null,"abstract":"\u0000 The heat transfer during pool boiling incorporates a higher rate of heat dissipation capability at low-temperature differences. This technique is widely used in the nuclear industry for thermal management. In this study, the effect of tube diameter and length on critical heat flux (CHF) at atmospheric conditions in saturated water during pool boiling is analyzed. The tubes of SS 304 are kept in the vertical orientation. The diameter of the tubes ranges from 1.2 mm to 9 mm. The tube lengths varying from 50mm to 1000mm. It has been noted that tubes of smaller diameter show a greater magnitude of CHF for the given length. For a given diameter, a longer tube is found to have lower CHF than the ones having lesser length. The variation in the CHF magnitude is negligible for tubes with a diameter of more than 2.5 mm beyond a length of 200 mm. The location of occurrence of CHF is near the bottom end of the vertical tube. The study illustrates the behavior of bubble nucleation for various tube dimensions and heat fluxes. The inception and detachment of bubbles for different tubes are analyzed. The pool boiling regime is categorized and studied basing the behaviour of the incepted and departed bubble. A mathematical relation that empirically accounts for the effect of tube dimensions i.e. length and diameter on pool boiling CHF is proposed. The experimental CHF data obtained during pool boiling are tabulated towards contributing to the CHF databank.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87405604","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The steel boxes that contain radioactive solid wastes should be accurately identified and measured. On account of the uneven internal filling and radioactive distribution in the steel boxes that storing radioactive solid wastes, the non-destructive measurement analysis technique can be utilized to perform overall measurement of the wastes, which will not damage the steel boxes and produce secondary wastes. The γ-ray spectrum measurement analysis algorithm of radioactive solid waste steel boxes was set up by way of discrete treatment for γ-ray spectrum measurement analysis algorithm combined with the traditional CT principle in this paper, the progressive split voxel measurement method was applied to conduct discrete treatment for the radioactive solid waste steel boxes, with the influence of voxel interactions on detection efficiency was taken into consideration, the calibration model of detection efficiency was set up through Monte Carlo simulation combined with algebraic reconstruction technique, the inversion correction of efficiency for the physically coincident part of the detected voxel volume was achieved. The proposed algorithm was preliminarily applied on the analyze platform of steel boxes that storing FA-IV- type radioactive solid wastes, and the detection and verification on the standard radioactive source of known activities was also conducted, the findings revealed that the relative deviation between the reconstruction results and truth-value of radioactive activities in the steel boxes is less than 30%, which conformed to the expected results and the proposed algorithm can be popularized and applicated in the detection works in related to other.
{"title":"Study of γ-ray Spectrum Measurement Analysis Algorithm of Radioactive Solid Waste Steel Boxes","authors":"Yurong Li, Lixia He, Jing Wen, Jiewen Shao","doi":"10.1115/1.4063039","DOIUrl":"https://doi.org/10.1115/1.4063039","url":null,"abstract":"\u0000 The steel boxes that contain radioactive solid wastes should be accurately identified and measured. On account of the uneven internal filling and radioactive distribution in the steel boxes that storing radioactive solid wastes, the non-destructive measurement analysis technique can be utilized to perform overall measurement of the wastes, which will not damage the steel boxes and produce secondary wastes. The γ-ray spectrum measurement analysis algorithm of radioactive solid waste steel boxes was set up by way of discrete treatment for γ-ray spectrum measurement analysis algorithm combined with the traditional CT principle in this paper, the progressive split voxel measurement method was applied to conduct discrete treatment for the radioactive solid waste steel boxes, with the influence of voxel interactions on detection efficiency was taken into consideration, the calibration model of detection efficiency was set up through Monte Carlo simulation combined with algebraic reconstruction technique, the inversion correction of efficiency for the physically coincident part of the detected voxel volume was achieved. The proposed algorithm was preliminarily applied on the analyze platform of steel boxes that storing FA-IV- type radioactive solid wastes, and the detection and verification on the standard radioactive source of known activities was also conducted, the findings revealed that the relative deviation between the reconstruction results and truth-value of radioactive activities in the steel boxes is less than 30%, which conformed to the expected results and the proposed algorithm can be popularized and applicated in the detection works in related to other.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82303388","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Srivastava, Saikrishna Nadella, N. K. Maheshwari
A Molten Salt Natural Circulation Loop (MSNCL) has been setup to study the steady state, transient and stability characteristics of molten nitrate salt, mixture of Sodium Nitrate and Potassium Nitrate in 60:40 ratios by weight. Natural circulation experiments in a temperature range of 290 °C to 600 °C have been performed. A semi-analytical linear model was derived for stability analysis of rectangular natural circulation loops with conventional localized surface heating and cooling. The developed model includes the effect of wall thermal inertia, variable internal heat transfer coefficient, finite secondary side heat transfer coefficient at cooler and heat losses for predicting the stability map. These effects are incorporated considering their significant role in modelling high temperature molten salt based natural circulation systems. The developed model has been first validated with the experimental data of water loop available in literature. The developed model is then validated with the experimental data generated in MSNCL. Validation of in-house developed non-linear model has also been performed against the same experimental data. The comparison of both linear and non-linear stability analysis with the experimental data shows good agreement and articulate the importance of various parameters which have been included in the developed model.
{"title":"Linear And Non-Linear Stability Analysis Of Molten Salt Natural Circulation Loop","authors":"A. Srivastava, Saikrishna Nadella, N. K. Maheshwari","doi":"10.1115/1.4063040","DOIUrl":"https://doi.org/10.1115/1.4063040","url":null,"abstract":"\u0000 A Molten Salt Natural Circulation Loop (MSNCL) has been setup to study the steady state, transient and stability characteristics of molten nitrate salt, mixture of Sodium Nitrate and Potassium Nitrate in 60:40 ratios by weight. Natural circulation experiments in a temperature range of 290 °C to 600 °C have been performed. A semi-analytical linear model was derived for stability analysis of rectangular natural circulation loops with conventional localized surface heating and cooling. The developed model includes the effect of wall thermal inertia, variable internal heat transfer coefficient, finite secondary side heat transfer coefficient at cooler and heat losses for predicting the stability map. These effects are incorporated considering their significant role in modelling high temperature molten salt based natural circulation systems. The developed model has been first validated with the experimental data of water loop available in literature. The developed model is then validated with the experimental data generated in MSNCL. Validation of in-house developed non-linear model has also been performed against the same experimental data. The comparison of both linear and non-linear stability analysis with the experimental data shows good agreement and articulate the importance of various parameters which have been included in the developed model.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74387250","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Le Mei, Xiaochun Zhang, Junbao Zhang, Changlei Shao, Jialei Zhu, Ran Huang, Chongzhi Wu
In order to realize the steel liner underwater repairing of the spent fuel pool of the third generation nuclear power plant, the laser welding process tests were carried out step by step in three environments: air, shallow water and simulating-repairing of the spent fuel pool floor(high-pressure condition). Through the process optimization, the high-quality forming of the underwater laser welding of duplex stainless steel was realized, and the underwater local dry laser welding process suitable for the spent fuel pool floor of nuclear power plant was developed. The results of nondestructive testing (including visual testing, liquid penetrant testing, ultrasonic testing and radiographic testing) of welding test pieces under three environments were qualified, and the test results of properties (including tensile, impact, bending, intergranular corrosion, ferrite content) meet the standard requirements. The underwater weld performance is similar to that in the air environment, and the weld quality meets the requirements of the spent fuel pool construction standard, laying a technical foundation for the application of the spent fuel pool underwater repairing.
{"title":"Study On the Step by Step Process and Performance of Laser Welding for the Spent Fuel Pool Floor","authors":"Le Mei, Xiaochun Zhang, Junbao Zhang, Changlei Shao, Jialei Zhu, Ran Huang, Chongzhi Wu","doi":"10.1115/1.4063008","DOIUrl":"https://doi.org/10.1115/1.4063008","url":null,"abstract":"\u0000 In order to realize the steel liner underwater repairing of the spent fuel pool of the third generation nuclear power plant, the laser welding process tests were carried out step by step in three environments: air, shallow water and simulating-repairing of the spent fuel pool floor(high-pressure condition). Through the process optimization, the high-quality forming of the underwater laser welding of duplex stainless steel was realized, and the underwater local dry laser welding process suitable for the spent fuel pool floor of nuclear power plant was developed. The results of nondestructive testing (including visual testing, liquid penetrant testing, ultrasonic testing and radiographic testing) of welding test pieces under three environments were qualified, and the test results of properties (including tensile, impact, bending, intergranular corrosion, ferrite content) meet the standard requirements. The underwater weld performance is similar to that in the air environment, and the weld quality meets the requirements of the spent fuel pool construction standard, laying a technical foundation for the application of the spent fuel pool underwater repairing.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82441959","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In Europe the design extension conditions (DEC) were introduced after the Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The objective of the study is to determine available elapsed time before core uncovery and needed DEC safety features for total loss of all feedwater (TLOFW) in a two-loop pressurized water reactor. RELAP5/MOD3.3 computer code has been used for calculations. The initiating event for TLOFW are multiple failures in which, besides the loss of main feedwater also the auxiliary feedwater is lost. The scenarios without DEC safety features and the scenarios with DEC safety features assumed have been simulated. The results showed that after TLOFW event initiation it is very important to trip the reactor as soon as possible. In case of loss of offsite power the reactor coolant pumps stop and the reactor very quickly trips on low reactor coolant pump flow. When normal operation systems are assumed the reactor trip occurs on low-low steam generator narrow level few tens of seconds after accident initiation, resulting in less time available before core uncovery occurence. The results for TLOFW scenarios with normal operation systems and DEC safety featured assumed demonstrated that secondary side bleed and feed can prevent core uncovery in case when no operator actions are credited before 30 minute. When primary side bleed and feed is used, less time is available for operator actions.
{"title":"Total Loss of Feedwater Analysis of PWR Using RELAP5","authors":"A. Prošek","doi":"10.1115/1.4063009","DOIUrl":"https://doi.org/10.1115/1.4063009","url":null,"abstract":"\u0000 In Europe the design extension conditions (DEC) were introduced after the Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The objective of the study is to determine available elapsed time before core uncovery and needed DEC safety features for total loss of all feedwater (TLOFW) in a two-loop pressurized water reactor. RELAP5/MOD3.3 computer code has been used for calculations. The initiating event for TLOFW are multiple failures in which, besides the loss of main feedwater also the auxiliary feedwater is lost. The scenarios without DEC safety features and the scenarios with DEC safety features assumed have been simulated.\u0000 The results showed that after TLOFW event initiation it is very important to trip the reactor as soon as possible. In case of loss of offsite power the reactor coolant pumps stop and the reactor very quickly trips on low reactor coolant pump flow. When normal operation systems are assumed the reactor trip occurs on low-low steam generator narrow level few tens of seconds after accident initiation, resulting in less time available before core uncovery occurence. The results for TLOFW scenarios with normal operation systems and DEC safety featured assumed demonstrated that secondary side bleed and feed can prevent core uncovery in case when no operator actions are credited before 30 minute. When primary side bleed and feed is used, less time is available for operator actions.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82871641","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The heavy water reactor concept Teplator is a pressure channel type reactor with independent systems for the primary coolant and the moderator. The present study analyses the low-pressure moderator cooling system of Teplator during full-power operation. The moderator is heated from neutron thermalization, gamma rays absorption, fission product decay and decay of activation products. Additionally, heat transfer from the coolant channels has to be taken in the analyses of the moderator cooling system. Preliminary thermal-hydraulic analyses of the cooling system are supplemented by CFD simulations of heat and fluid flow in the moderator's vessel, emphasizing flow-type regimes. Results from CFD simulations showed that the buoyancy-dominated flow (case MF-22) resulted in a higher thermal stratification and high moderator temperature close to the upper plate of the moderator vessel. The inertia-dominated flow regime MF-90 resulted in good mixing of the moderator and a low thermal stratification in the vessel. Finally, the mid-mass flow rate regime MF-45 was identified as a transitional region from a buoyancy-dominated to a mix-type regime.
{"title":"CFD Calculations of Moderator Heat and Fluid Flow of Small Modular Heavy Water Reactor","authors":"T. Kořínek, R. Škoda, M. Lovecký, O. Burian","doi":"10.1115/1.4063007","DOIUrl":"https://doi.org/10.1115/1.4063007","url":null,"abstract":"\u0000 The heavy water reactor concept Teplator is a pressure channel type reactor with independent systems for the primary coolant and the moderator. The present study analyses the low-pressure moderator cooling system of Teplator during full-power operation. The moderator is heated from neutron thermalization, gamma rays absorption, fission product decay and decay of activation products. Additionally, heat transfer from the coolant channels has to be taken in the analyses of the moderator cooling system. Preliminary thermal-hydraulic analyses of the cooling system are supplemented by CFD simulations of heat and fluid flow in the moderator's vessel, emphasizing flow-type regimes. Results from CFD simulations showed that the buoyancy-dominated flow (case MF-22) resulted in a higher thermal stratification and high moderator temperature close to the upper plate of the moderator vessel. The inertia-dominated flow regime MF-90 resulted in good mixing of the moderator and a low thermal stratification in the vessel. Finally, the mid-mass flow rate regime MF-45 was identified as a transitional region from a buoyancy-dominated to a mix-type regime.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-07-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81450350","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Rhodium is one of the scarcest, most valuable, and useful platinum group metals, a strategically important material relied on heavily by automotive and electronics industries. The limited finite natural sources of Rh and exponentially increasing demands on these supplies mean that new sources are being sought to stabilise supplies and prices. Spent nuclear fuel (SNF) contains a significant quantity of Rh, though methods to recover this are purely conceptual at this point, due to the differing chemistry between SNF reprocessing and the methods used to recycle natural Rh. During SNF reprocessing, Rh partitions between aqueous nitric acid streams, where its speciation is complex, and insoluble fission product waste streams. Various techniques have been investigated for Rh recovery during SNF reprocessing for over 50 years, including solvent extraction, ion exchange, precipitation, and electrochemical methods, with tuneable approaches such as impregnated composites and ionic liquids receiving the most attention recently, assisted by more the comprehensive understanding of Rh speciation in nitric acid developed recently. The quantitative recovery of Rh within the SNF reprocessing ecosystem has remained elusive thus far, and as such, this review discusses the recent developments within the field, and strategies that could be applied to maximise the recovery of Rh from SNF.
{"title":"A Review of Opportunities and Methods for Recovery of Rhodium from Spent Nuclear Fuel during Reprocessing","authors":"B. Hodgson, J. Turner, Alistair F. Holdsworth","doi":"10.3390/jne4030034","DOIUrl":"https://doi.org/10.3390/jne4030034","url":null,"abstract":"Rhodium is one of the scarcest, most valuable, and useful platinum group metals, a strategically important material relied on heavily by automotive and electronics industries. The limited finite natural sources of Rh and exponentially increasing demands on these supplies mean that new sources are being sought to stabilise supplies and prices. Spent nuclear fuel (SNF) contains a significant quantity of Rh, though methods to recover this are purely conceptual at this point, due to the differing chemistry between SNF reprocessing and the methods used to recycle natural Rh. During SNF reprocessing, Rh partitions between aqueous nitric acid streams, where its speciation is complex, and insoluble fission product waste streams. Various techniques have been investigated for Rh recovery during SNF reprocessing for over 50 years, including solvent extraction, ion exchange, precipitation, and electrochemical methods, with tuneable approaches such as impregnated composites and ionic liquids receiving the most attention recently, assisted by more the comprehensive understanding of Rh speciation in nitric acid developed recently. The quantitative recovery of Rh within the SNF reprocessing ecosystem has remained elusive thus far, and as such, this review discusses the recent developments within the field, and strategies that could be applied to maximise the recovery of Rh from SNF.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.4,"publicationDate":"2023-07-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79205071","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：