The distribution of hydrogen inside the containment is a key issue in assessing the evolution of the postulated accident. For safety analysis and codes validation purposes, a large scale comprehensive test facility has been built to investigate the containment thermal-hydraulic characteristics under accident conditions. In this paper, a test was performed to experimentally investigate the distribution of the hydrogen inside the containment and the influence of the external cooling on gas mixing and stratification. The paper presents the experimental results of the integral test performed in this facility. During the experiments, helium was used to simulate hydrogen. Helium and steam are released together and allowed to take additional time to form a relatively stable stratification, then followed by external cooling. The initial pressure of the experiments is around 0.1MPa(a) and the initial Froude number is around 333. The results showed that a helium-enriched stratification emerged in the upper containment due to the density difference after the injection. External cooling caused condensation and intense convective flow. As a result, an overall increase in helium concentration was observed with a decrease in concentration gradient.
{"title":"Experimental Investigation on Gas Mixing and Stratification in Containment Influenced by External Cooling","authors":"Li Ying, W. Xue, Cao Xuewu, L. Tong","doi":"10.1115/icone2020-16430","DOIUrl":"https://doi.org/10.1115/icone2020-16430","url":null,"abstract":"\u0000 The distribution of hydrogen inside the containment is a key issue in assessing the evolution of the postulated accident. For safety analysis and codes validation purposes, a large scale comprehensive test facility has been built to investigate the containment thermal-hydraulic characteristics under accident conditions. In this paper, a test was performed to experimentally investigate the distribution of the hydrogen inside the containment and the influence of the external cooling on gas mixing and stratification. The paper presents the experimental results of the integral test performed in this facility. During the experiments, helium was used to simulate hydrogen. Helium and steam are released together and allowed to take additional time to form a relatively stable stratification, then followed by external cooling. The initial pressure of the experiments is around 0.1MPa(a) and the initial Froude number is around 333. The results showed that a helium-enriched stratification emerged in the upper containment due to the density difference after the injection. External cooling caused condensation and intense convective flow. As a result, an overall increase in helium concentration was observed with a decrease in concentration gradient.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"53 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115297973","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}
Sun Jianchuang, M. Ding, Zhengpeng Mi, Zhang Zhuohua
Direct contact condensation (DCC) is a common physical phenomenon appearing in the nuclear power plants and other industrial applications. The current research on DCC focuses on steam-water counterflow or jet flow under forced flow conditions. For some natural-circulation passive safety systems in floating nuclear power plants, the heated section is connected with the heat sink by the horizontal pipes. The heat sink is usually the ocean. In such natural circulation systems (NCSs), the steam produced in the heated section will inevitably contact the subcooled water in the ocean and result in DCC event. In this paper, the fundamental characteristics of two-phase flow were described in detail. In addition, the influences of the subcooled water temperature on the flow rate, outlet temperature, steam bubble behaviors, and pressure surge were emphatically discussed. The experimental results revealed that the subcooled water in the water tank could reversely flow into the pipe, which was able to cause condensation induced water hammer (CIWH) event and flow oscillation. Further research showed that the subcooled water was evidently prevented from reversely flowing into the pipe with the increase in the subcooled water temperature. The position where the DCC event occurs transferred from the pipe to the water tank, and the pressure peak obviously reduced, even disappear when the subcooled water temperature is larger than 61 °C.
{"title":"Experimental Investigation on Temperature Characteristics of Direct Contact Condensation in a Natural Circulation System","authors":"Sun Jianchuang, M. Ding, Zhengpeng Mi, Zhang Zhuohua","doi":"10.1115/icone2020-16257","DOIUrl":"https://doi.org/10.1115/icone2020-16257","url":null,"abstract":"\u0000 Direct contact condensation (DCC) is a common physical phenomenon appearing in the nuclear power plants and other industrial applications. The current research on DCC focuses on steam-water counterflow or jet flow under forced flow conditions. For some natural-circulation passive safety systems in floating nuclear power plants, the heated section is connected with the heat sink by the horizontal pipes. The heat sink is usually the ocean. In such natural circulation systems (NCSs), the steam produced in the heated section will inevitably contact the subcooled water in the ocean and result in DCC event. In this paper, the fundamental characteristics of two-phase flow were described in detail. In addition, the influences of the subcooled water temperature on the flow rate, outlet temperature, steam bubble behaviors, and pressure surge were emphatically discussed. The experimental results revealed that the subcooled water in the water tank could reversely flow into the pipe, which was able to cause condensation induced water hammer (CIWH) event and flow oscillation. Further research showed that the subcooled water was evidently prevented from reversely flowing into the pipe with the increase in the subcooled water temperature. The position where the DCC event occurs transferred from the pipe to the water tank, and the pressure peak obviously reduced, even disappear when the subcooled water temperature is larger than 61 °C.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"45 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116280518","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}
To investigate the heat transfer characteristics of the chrome-plated tube and hope it could be used on the internal heat exchanger of passive containment cooling system (PCS), an experimental investigation has been conducted. In this experiment, a series of steam condensation experiments are performed under pure steam and steam-air mixed conditions over chrome-plated tube for a variety of chromium coating thickness (1μm and 10μm), total pressure, air mass fraction and wall subcooling. Condensation heat transfer coefficient was obtained for the total pressure ranging from 0.2 MPa to 0.4 MPa, air mass fraction ranging from 0.10 to 0.71, and wall subcooling from 10°C to 70°C. Moreover, the designed visualization experimental device makes the experimental phenomenon can be directly observed through the observation window. Under the pure steam condition, the results show that droplet condensation and filmwise condensation is co-existed on both two kinds of chrome-plated tubes, the chrome coating thickness of 10μm tube shows better heat transfer ability. Under the steam-air mixed condition, the condensation heat transfer coefficient of both two kinds of tubes increases with total pressure, and decrease with the air mass fraction and wall subcooling, while the influence of chrome coating thickness on heat transfer is no longer noticeable. The results also indicate that the thickness of the chromium coating will affect the surface microstructure of the chrome-plated tube and then affect the heat transfer ability of the chrome-plated tube.
{"title":"Experimental Study of Pure Steam and Steam-Air Mixture Condensation on a Vertical Chrome-Plated Tube","authors":"Z. Niu, Guangming Fan, Jie Cheng, Wei Li","doi":"10.1115/icone2020-16204","DOIUrl":"https://doi.org/10.1115/icone2020-16204","url":null,"abstract":"\u0000 To investigate the heat transfer characteristics of the chrome-plated tube and hope it could be used on the internal heat exchanger of passive containment cooling system (PCS), an experimental investigation has been conducted. In this experiment, a series of steam condensation experiments are performed under pure steam and steam-air mixed conditions over chrome-plated tube for a variety of chromium coating thickness (1μm and 10μm), total pressure, air mass fraction and wall subcooling. Condensation heat transfer coefficient was obtained for the total pressure ranging from 0.2 MPa to 0.4 MPa, air mass fraction ranging from 0.10 to 0.71, and wall subcooling from 10°C to 70°C. Moreover, the designed visualization experimental device makes the experimental phenomenon can be directly observed through the observation window. Under the pure steam condition, the results show that droplet condensation and filmwise condensation is co-existed on both two kinds of chrome-plated tubes, the chrome coating thickness of 10μm tube shows better heat transfer ability. Under the steam-air mixed condition, the condensation heat transfer coefficient of both two kinds of tubes increases with total pressure, and decrease with the air mass fraction and wall subcooling, while the influence of chrome coating thickness on heat transfer is no longer noticeable. The results also indicate that the thickness of the chromium coating will affect the surface microstructure of the chrome-plated tube and then affect the heat transfer ability of the chrome-plated tube.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"82 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126172300","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}
Anupreethi Balajiranganathan, Anurag Gupta, Umasankari Kannan, A. Tiwari
A solution to optimization of in-core detectors placement for Advanced Heavy Water Reactor (AHWR) has been attempted. AHWR houses in-core detector units with Self-Powered Neutron Detectors (SPND) distributed axially and their measurement serves as an input to Online Flux Mapping System (OFMS) to monitor the three-dimensional neutron flux distribution. There is a requirement of placing these in-core detectors at optimum locations to retrieve maximum information about the reactor while keeping their number to the minimum. This paper attempts to optimize SPND placement through the application of Bond Energy Algorithm (BEA), a clustering technique which groups the SPNDs based on correlation. This works on the concept of grouping strongly correlated SPNDs into blocks and choosing one SPND from each block as the optimal location. The higher the uncorrelation among optimal SPNDs, the higher the independent information retrieved about the actual configuration of the reactor. It can be inferred from this work that the number and location of SPNDs are highly dependent on the initial set of SPND locations and the correlation threshold. It can be seen that as the correlation threshold increases, the number of optimal locations increases. The obtained optimal locations have been validated for various operational reactor configurations using different Flux Mapping Algorithms (FMA).
{"title":"Optimization of In-Core Detector Locations in AHWR Using Bond Energy Algorithm","authors":"Anupreethi Balajiranganathan, Anurag Gupta, Umasankari Kannan, A. Tiwari","doi":"10.1115/icone2020-16480","DOIUrl":"https://doi.org/10.1115/icone2020-16480","url":null,"abstract":"\u0000 A solution to optimization of in-core detectors placement for Advanced Heavy Water Reactor (AHWR) has been attempted. AHWR houses in-core detector units with Self-Powered Neutron Detectors (SPND) distributed axially and their measurement serves as an input to Online Flux Mapping System (OFMS) to monitor the three-dimensional neutron flux distribution. There is a requirement of placing these in-core detectors at optimum locations to retrieve maximum information about the reactor while keeping their number to the minimum. This paper attempts to optimize SPND placement through the application of Bond Energy Algorithm (BEA), a clustering technique which groups the SPNDs based on correlation. This works on the concept of grouping strongly correlated SPNDs into blocks and choosing one SPND from each block as the optimal location. The higher the uncorrelation among optimal SPNDs, the higher the independent information retrieved about the actual configuration of the reactor. It can be inferred from this work that the number and location of SPNDs are highly dependent on the initial set of SPND locations and the correlation threshold. It can be seen that as the correlation threshold increases, the number of optimal locations increases. The obtained optimal locations have been validated for various operational reactor configurations using different Flux Mapping Algorithms (FMA).","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"20 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125347272","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}
Nuclear safety and security is getting more and more attention all over the world. Vehicles in and out of nuclear facilities need to be strictly inspected to prevent carrying of nuclear materials, explosives or other dangerous items. A new-type vehicle inspection system — 60Co gantry-movable dual-projection digital radiography inspection system is developed in Tsinghua University under the support of China’s Nuclear Energy Development Project, which uses two 60Co as radiation sources. The radiation sources are arranged at the bottom and side of the vehicles to be detected and the ionization chamber detectors are set in the two side of the gantry correspondingly. With moving of the sources and gantry synchronously, the system can obtain both side-view and upward-view images of the vehicles simultaneously [1]. However, there is a problem that the scattered rays from source of one projection can enter the detector array of another projection to form an interference signal. Experiments show that this kind of scattering noise can account for 20%, resulting in blurring and even artifacts, especially in thicker areas of the image. This problem needs to be solved urgently. According to the characteristics of the Compton scattering effect, it is inferred that there is a certain non-linear mapping relationship between the scattering distribution of the detector array of one projection and the mass distribution in the plane of the other projection. This paper attempts to use BP neural network to learn this mapping relationship to quantitatively remove this kind of scattering noise. The results show that this method has certain effects on the removal of artifacts and blur caused by scattering. This method has the advantage of being fast and more targeted, compared with traditional image post-processing methods.
{"title":"Study on Scattering Correction of the 60Co Gantry-Movable Dual-Projection Digital Radiography Inspection System","authors":"Minzi Ni, Guang-sheng Li, Zhentao Wang","doi":"10.1115/icone2020-16320","DOIUrl":"https://doi.org/10.1115/icone2020-16320","url":null,"abstract":"\u0000 Nuclear safety and security is getting more and more attention all over the world. Vehicles in and out of nuclear facilities need to be strictly inspected to prevent carrying of nuclear materials, explosives or other dangerous items. A new-type vehicle inspection system — 60Co gantry-movable dual-projection digital radiography inspection system is developed in Tsinghua University under the support of China’s Nuclear Energy Development Project, which uses two 60Co as radiation sources. The radiation sources are arranged at the bottom and side of the vehicles to be detected and the ionization chamber detectors are set in the two side of the gantry correspondingly. With moving of the sources and gantry synchronously, the system can obtain both side-view and upward-view images of the vehicles simultaneously [1]. However, there is a problem that the scattered rays from source of one projection can enter the detector array of another projection to form an interference signal. Experiments show that this kind of scattering noise can account for 20%, resulting in blurring and even artifacts, especially in thicker areas of the image. This problem needs to be solved urgently.\u0000 According to the characteristics of the Compton scattering effect, it is inferred that there is a certain non-linear mapping relationship between the scattering distribution of the detector array of one projection and the mass distribution in the plane of the other projection. This paper attempts to use BP neural network to learn this mapping relationship to quantitatively remove this kind of scattering noise. The results show that this method has certain effects on the removal of artifacts and blur caused by scattering. This method has the advantage of being fast and more targeted, compared with traditional image post-processing methods.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"28 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114717690","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 an over pressure accident, one or more pressurizer safety (or relief) valves will open due to the rapid pressure rise process. Once the safety (or relief) valves are open, the liquid seal will be discharged, and this will generate great discharge force to the downstream pipes. Multi-level protection is chosen using pressurizer safety (or relief) valves with different setpoint in most of Nuclear Power Plant, especially in the self-designed Generation-III Nuclear Power Plants. As the over pressure accident progresses, one or more safety (or relief) valves will be open. The downstream pipes will experience one or more times of impacts, which will influence the arrangement of the pipes. The whole discharge process is very complex, and the key influence factors are the pressure rise rate, safety (or relief) valve opening time, liquid seal temperature and volume, and the arrangement of the downstream discharge pipes. In present paper, liquid seal discharge process in an over pressure accident is studied. The pressure rise rate is so fast that three safety (or relief) valves will open one after another, which will generate three impacts on the downstream discharge pipes. It is found that for a specific design of Nuclear Power Plant, well design of the safety (or relief) valve setpoint is very important to the discharge force analysis results.
{"title":"Study on the Liquid Seal Discharge Process in an Over-Pressurized Accident","authors":"Dan Wu, Deng Jian, Du Sijia, Qian Libo","doi":"10.1115/icone2020-16786","DOIUrl":"https://doi.org/10.1115/icone2020-16786","url":null,"abstract":"\u0000 In an over pressure accident, one or more pressurizer safety (or relief) valves will open due to the rapid pressure rise process. Once the safety (or relief) valves are open, the liquid seal will be discharged, and this will generate great discharge force to the downstream pipes. Multi-level protection is chosen using pressurizer safety (or relief) valves with different setpoint in most of Nuclear Power Plant, especially in the self-designed Generation-III Nuclear Power Plants. As the over pressure accident progresses, one or more safety (or relief) valves will be open. The downstream pipes will experience one or more times of impacts, which will influence the arrangement of the pipes. The whole discharge process is very complex, and the key influence factors are the pressure rise rate, safety (or relief) valve opening time, liquid seal temperature and volume, and the arrangement of the downstream discharge pipes. In present paper, liquid seal discharge process in an over pressure accident is studied. The pressure rise rate is so fast that three safety (or relief) valves will open one after another, which will generate three impacts on the downstream discharge pipes. It is found that for a specific design of Nuclear Power Plant, well design of the safety (or relief) valve setpoint is very important to the discharge force analysis results.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"15 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127985023","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}
Pei-Syuan Ruan, Shao-Wen Chen, Min-Song Lin, Jin-Der Lee, Jong-Rong Wang
This paper presents the experimental results and analyses of the structure velocity of air-water two-phase flow in a 3 × 3 rod bundle channel. A total of 56 flow conditions were tested and investigated for rod-gap, sub-channel, rod-wall and global regions of rod bundle geometry. The experimental tests were carried out under bubbly and cap-bubbly flow regimes with superficial gas and liquid velocities of 0–1 m/s and 1–1.7 m/s, respectively. The conductivity probes were set at different heights to measure the global and local void fractions. The structure velocity of air-water two-phase flow is the average bubble velocity calculated by the method in this study. The structure velocity were determined by utilizing the cross-correlation technique to analyze the time lags of the bubbles passing through the conductivity probes. The results of this study indicated that the structure velocity may increase with increasing superficial gas and liquid velocities. In low superficial gas velocity region, the structure velocity may first slightly increase and follow by a sudden jump which appear in most regions. After the sudden jump, the structure velocity may keep increasing mildly. The present structure velocities have been compared with the area-averaged gas velocities predicted by the drift flux model, and it appears that most structure velocities show a good agreement with the averaged gas velocities from the drift flux model after the jump.
{"title":"Investigation of the Structure Velocity in a 3x3 Rod Bundle Under Bubbly and Cap-Bubbly Flow Regimes","authors":"Pei-Syuan Ruan, Shao-Wen Chen, Min-Song Lin, Jin-Der Lee, Jong-Rong Wang","doi":"10.1115/icone2020-16946","DOIUrl":"https://doi.org/10.1115/icone2020-16946","url":null,"abstract":"\u0000 This paper presents the experimental results and analyses of the structure velocity of air-water two-phase flow in a 3 × 3 rod bundle channel. A total of 56 flow conditions were tested and investigated for rod-gap, sub-channel, rod-wall and global regions of rod bundle geometry. The experimental tests were carried out under bubbly and cap-bubbly flow regimes with superficial gas and liquid velocities of 0–1 m/s and 1–1.7 m/s, respectively. The conductivity probes were set at different heights to measure the global and local void fractions. The structure velocity of air-water two-phase flow is the average bubble velocity calculated by the method in this study. The structure velocity were determined by utilizing the cross-correlation technique to analyze the time lags of the bubbles passing through the conductivity probes. The results of this study indicated that the structure velocity may increase with increasing superficial gas and liquid velocities. In low superficial gas velocity region, the structure velocity may first slightly increase and follow by a sudden jump which appear in most regions. After the sudden jump, the structure velocity may keep increasing mildly. The present structure velocities have been compared with the area-averaged gas velocities predicted by the drift flux model, and it appears that most structure velocities show a good agreement with the averaged gas velocities from the drift flux model after the jump.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129985171","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 Transient Reactor Test (TREAT) Facility at Idaho National Laboratory (INL) started testing new fuels and reactor technologies once again in 2018 and new experiments and tests are currently being designed like for example the water loop “TREAT Water Environment Recirculating Loop” (TWERL). During the design of such experiments, the designer must assess how close the experiment reproduces the physics (and other important phenomena) happening during a transient of interest compared to the full-size reactor the experiment attempts representing. Traditionally, to assess this “representativity” of the experiment, scaling theory involving expert judgment is needed. This paper presents a step towards a systematic modeling and simulation (M&S) informed methodology for experiment design. The new methodology compares a model of the full system and a model of the mock-up facility that are subject to the same perturbations. In this way, the “overlap” of the perturbed experiment and full-size facility model outputs can be analyzed and the “representativity” of the experiment determined. The paper presents a RELAP5-3D analysis, where TWERL LOCA calculations are compared to prototypic PWR LOCA calculations with respect to representativity. To inform the design of the TWERL experiments, i.e. to find the most “representative” configuration for the TWERL loop, different design parameters for TWERL have been optimized in the study.
{"title":"Representativity Analysis Applied to TREAT Water Loop LOCA Experiment Design","authors":"A. Epiney, N. Woolstenhulme","doi":"10.1115/icone2020-16914","DOIUrl":"https://doi.org/10.1115/icone2020-16914","url":null,"abstract":"\u0000 The Transient Reactor Test (TREAT) Facility at Idaho National Laboratory (INL) started testing new fuels and reactor technologies once again in 2018 and new experiments and tests are currently being designed like for example the water loop “TREAT Water Environment Recirculating Loop” (TWERL). During the design of such experiments, the designer must assess how close the experiment reproduces the physics (and other important phenomena) happening during a transient of interest compared to the full-size reactor the experiment attempts representing. Traditionally, to assess this “representativity” of the experiment, scaling theory involving expert judgment is needed. This paper presents a step towards a systematic modeling and simulation (M&S) informed methodology for experiment design. The new methodology compares a model of the full system and a model of the mock-up facility that are subject to the same perturbations. In this way, the “overlap” of the perturbed experiment and full-size facility model outputs can be analyzed and the “representativity” of the experiment determined. The paper presents a RELAP5-3D analysis, where TWERL LOCA calculations are compared to prototypic PWR LOCA calculations with respect to representativity. To inform the design of the TWERL experiments, i.e. to find the most “representative” configuration for the TWERL loop, different design parameters for TWERL have been optimized in the study.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"33 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132085407","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 sensitivity analysis of the dynamical response of reactor coolant system to the input parameters is an important precondition for the design optimization. In this paper, the sensitivity of the dynamical loads at the nozzles of the equipment under seismic conditions is analyzed with an integrated platform called OPTIMUS, taking the stiffness of the dampers in the steam generator and the main pump as the input variables. The key parameters of the reactor system are usually different from the design value due to the calculation error, random and other uncontrollable errors in the manufacturing process and installation process. In a nuclear power project, the measured stiffness values of the dampers on the steam generator and the main pump in the manufacturer are deviated from the requirements in the equipment specification, and it is necessary to evaluate the influence of the deviation on the dynamical response analysis of the reactor system. According to the traditional method, it is necessary to establish the models of the reactor coolant system for nonlinear analysis according to the different stiffness of the dampers, and then the calculation results are compared by EXCEL. In this paper, the sensitivity analysis of output parameters which are the loads at the nozzles of the equipment to the input parameters which are the stiffness of the dampers on the steam generator and pump is realized by OPTIMUS, which is a kind of integration platform. Not only can ANSYS simulation calculations be carried out automatically on the OPTIMUS, but also the output data can be processed rapidly automatically, and the influence of manufacturing deviation of the stiffness of the dampers on the dynamical response of the reactor coolant system can be analyzed quantitatively in the above-mentioned problems, and the data support is provided for the determination of the design variables for subsequent optimization analysis.
{"title":"Sensitivity Analysis for Dynamical Response of Reactor Coolant System Based on OPTIMUS","authors":"Yuan Yanli, Ye Xianhui, Li Lijuan, Yuan Feng","doi":"10.1115/icone2020-16485","DOIUrl":"https://doi.org/10.1115/icone2020-16485","url":null,"abstract":"\u0000 The sensitivity analysis of the dynamical response of reactor coolant system to the input parameters is an important precondition for the design optimization. In this paper, the sensitivity of the dynamical loads at the nozzles of the equipment under seismic conditions is analyzed with an integrated platform called OPTIMUS, taking the stiffness of the dampers in the steam generator and the main pump as the input variables. The key parameters of the reactor system are usually different from the design value due to the calculation error, random and other uncontrollable errors in the manufacturing process and installation process. In a nuclear power project, the measured stiffness values of the dampers on the steam generator and the main pump in the manufacturer are deviated from the requirements in the equipment specification, and it is necessary to evaluate the influence of the deviation on the dynamical response analysis of the reactor system.\u0000 According to the traditional method, it is necessary to establish the models of the reactor coolant system for nonlinear analysis according to the different stiffness of the dampers, and then the calculation results are compared by EXCEL. In this paper, the sensitivity analysis of output parameters which are the loads at the nozzles of the equipment to the input parameters which are the stiffness of the dampers on the steam generator and pump is realized by OPTIMUS, which is a kind of integration platform. Not only can ANSYS simulation calculations be carried out automatically on the OPTIMUS, but also the output data can be processed rapidly automatically, and the influence of manufacturing deviation of the stiffness of the dampers on the dynamical response of the reactor coolant system can be analyzed quantitatively in the above-mentioned problems, and the data support is provided for the determination of the design variables for subsequent optimization analysis.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"6 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121736330","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}
Guo Chao, Deng Jian, Cai Rong, Ma Yugao, L. Lili, Zhang Yu-hao, Lv Siyu
According to spray cooling characteristics for failure of the cooling system in the spent fuel pool, we establish a model for analyzing spray cooling of a single spent fuel rod and develop a code named SFSC which is used to calculate the thermal-hydraulic characteristics of film on the spent fuel rod under the condition of spray cooling. The results obtained from SFSC consist of film thickness, film temperature, film evaporation, fuel pellet temperature of spent fuel rod and cladding temperature of spent fuel rod at different height of spent fuel rod are shown in this paper. At the same time, we establish same model by RELAP and compare results calculated by RELAP with that calculated by SFSC. The good agreement of two results demonstrates that SFSC is effective. SFSC fills in the blanks of thermal-hydraulic calculation for the spent fuel rod under the condition of spray cooling.
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