Pressure reducing valve (PRV) located before the start-up vessel (SUV) is an essential component that decreases the pressure and temperature of the supercritical state steam by using spray water before it flows into the SUV. The PRV is kept closed during normal operation but opened during start-up and shutdown events, which could initiate thermal fatigue defects due to significant temperature changes. In addition to the thermal shock and internal pressure, system bending and torsional moments may be imparted on the PRV, threatening its integrity. To reinforce these concerns, cracks on the inside surfaces of the PRV have often been reported during planned maintenance activities in nuclear power plants. This research aims at analysing cyclic plasticity of the PRV subjected to cyclic moments, thermal and pressure loadings by means of an advanced direct numerical technique known as the Linear Matching Method (LMM). The cyclic moments are comprised of in-plane and out-of-plane bending, and torsion which are applied to an inlet branch pipe of the PRV. The cyclic thermal load is obtained from the transient heat transfer analysis using real operational data. Two different pressures, which are high and low pressures, are applied to internal surfaces of the PRV body and outlet pipe respectively. The analysed results construct a structural response boundary such as a shakedown limit boundary. The obtained structural response boundary is validated by full cyclic incremental analysis referred to as the step-by-step analysis. The analysed results have demonstrated that the plastic collapse limit is identical to the shakedown limit. Moreover, the results provide engineers with a safe load bearing capacity domain which otherwise requires evaluating structural integrity of the PRV subjected to the complicated cyclic loading condition using detailed assessments and analyses.
{"title":"Investigating Structural Response of Pressure Reducing Valve of Supercritical Steam Generator System Under Cyclic Moments, Thermal Transient, and Pressure Loadings","authors":"N. Cho, P. Bansal, A. Hurst","doi":"10.1115/icone2020-16426","DOIUrl":"https://doi.org/10.1115/icone2020-16426","url":null,"abstract":"\u0000 Pressure reducing valve (PRV) located before the start-up vessel (SUV) is an essential component that decreases the pressure and temperature of the supercritical state steam by using spray water before it flows into the SUV. The PRV is kept closed during normal operation but opened during start-up and shutdown events, which could initiate thermal fatigue defects due to significant temperature changes. In addition to the thermal shock and internal pressure, system bending and torsional moments may be imparted on the PRV, threatening its integrity. To reinforce these concerns, cracks on the inside surfaces of the PRV have often been reported during planned maintenance activities in nuclear power plants. This research aims at analysing cyclic plasticity of the PRV subjected to cyclic moments, thermal and pressure loadings by means of an advanced direct numerical technique known as the Linear Matching Method (LMM). The cyclic moments are comprised of in-plane and out-of-plane bending, and torsion which are applied to an inlet branch pipe of the PRV. The cyclic thermal load is obtained from the transient heat transfer analysis using real operational data. Two different pressures, which are high and low pressures, are applied to internal surfaces of the PRV body and outlet pipe respectively. The analysed results construct a structural response boundary such as a shakedown limit boundary. The obtained structural response boundary is validated by full cyclic incremental analysis referred to as the step-by-step analysis. The analysed results have demonstrated that the plastic collapse limit is identical to the shakedown limit. Moreover, the results provide engineers with a safe load bearing capacity domain which otherwise requires evaluating structural integrity of the PRV subjected to the complicated cyclic loading condition using detailed assessments and analyses.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73924829","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}
At the Hanford Waste Treatment and Immobilization Plant (WTP), various vessels are designed to be agitated with internal pulse jet mixers (PJMs) in order to provide a means of mixing with no moving parts local to the vessel. PJMs are operated by use of an applied vacuum to draw liquid in followed by motive air to force liquid out (while not completely discharging all the liquid). This continual operation results in mixing of the vessel contents. In off-normal conditions, PJMs may completely discharge resulting in air rapidly injected into the vessel (PJM overblow). An evaluation is complete to determine the statically applied transient Rayleigh-Plesset bubble loads resulting from PJM overblow on the vessel’s internal submerged structures. The low-frequency bubble loads on internal structures is determined via analysis of overblow test data, application of the Rayleigh-Plesset equation based on bubble pressure, PJM nozzle critical flow ratios, conservation of momentum, the relative equation of motion of a submerged non-fixed structure subject to both relative drag and relative acceleration, non-flow boundary conditions, use of a displacement-response spectra, and Hooke’s Law. This theoretical Rayleigh-Plesset bubble loads model accounts for various vessel and internal submerged structure designs and different operational states: PJM cavity pressure, liquid density, depth of submerged bubble, and both choked or non-choked flow through the PJM nozzle.
{"title":"Static Application of Transient Hydrodynamic Loads on Vessel Internal Structures As a Result of Pulse Jet Mixer Overblow: Low-Frequency Loads","authors":"Rafael Garcilazo, B. Fant, R. Blevins","doi":"10.1115/icone2020-17003","DOIUrl":"https://doi.org/10.1115/icone2020-17003","url":null,"abstract":"\u0000 At the Hanford Waste Treatment and Immobilization Plant (WTP), various vessels are designed to be agitated with internal pulse jet mixers (PJMs) in order to provide a means of mixing with no moving parts local to the vessel. PJMs are operated by use of an applied vacuum to draw liquid in followed by motive air to force liquid out (while not completely discharging all the liquid). This continual operation results in mixing of the vessel contents. In off-normal conditions, PJMs may completely discharge resulting in air rapidly injected into the vessel (PJM overblow).\u0000 An evaluation is complete to determine the statically applied transient Rayleigh-Plesset bubble loads resulting from PJM overblow on the vessel’s internal submerged structures. The low-frequency bubble loads on internal structures is determined via analysis of overblow test data, application of the Rayleigh-Plesset equation based on bubble pressure, PJM nozzle critical flow ratios, conservation of momentum, the relative equation of motion of a submerged non-fixed structure subject to both relative drag and relative acceleration, non-flow boundary conditions, use of a displacement-response spectra, and Hooke’s Law.\u0000 This theoretical Rayleigh-Plesset bubble loads model accounts for various vessel and internal submerged structure designs and different operational states: PJM cavity pressure, liquid density, depth of submerged bubble, and both choked or non-choked flow through the PJM nozzle.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74873385","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}
M. Ejiri, T. Kubota, Y. Soga, Nozomi Nishihara, N. Yanagida, Hiroomi Katou, T. Arashiba, Wataru Taniura
There are three types of hardening laws for evaluating welding residual stress with the finite element method (FEM): kinematic hardening law, isotropic hardening law, and combined hardening law that combine these. The purpose of this paper is to investigate which hardening law is more appropriate for the evaluation of welding residual stress of alloy 82. We first performed two types of welding tests: welding both ends of a plate, and welding the periphery of a disc. We then compared the results of mock-up welding tests with the analysis results of welding residual stress with the kinematic hardening law and combined hardening law. Both the kinematic hardening law and the combined hardening law showed a welding residual stress distribution close to the results of the mock-up welding tests, but the combined hardening law tended to be closer to the mock-up results. Therefore when it is necessary to confirm the welding residual stress of alloy 82, it is considered appropriate to apply the combined hardening law.
{"title":"Investigation of Hardening Law on Welding Residual Stress Analysis for Nickel Based Alloy 82 Weld Metal","authors":"M. Ejiri, T. Kubota, Y. Soga, Nozomi Nishihara, N. Yanagida, Hiroomi Katou, T. Arashiba, Wataru Taniura","doi":"10.1115/icone2020-16773","DOIUrl":"https://doi.org/10.1115/icone2020-16773","url":null,"abstract":"\u0000 There are three types of hardening laws for evaluating welding residual stress with the finite element method (FEM): kinematic hardening law, isotropic hardening law, and combined hardening law that combine these.\u0000 The purpose of this paper is to investigate which hardening law is more appropriate for the evaluation of welding residual stress of alloy 82. We first performed two types of welding tests: welding both ends of a plate, and welding the periphery of a disc. We then compared the results of mock-up welding tests with the analysis results of welding residual stress with the kinematic hardening law and combined hardening law.\u0000 Both the kinematic hardening law and the combined hardening law showed a welding residual stress distribution close to the results of the mock-up welding tests, but the combined hardening law tended to be closer to the mock-up results. Therefore when it is necessary to confirm the welding residual stress of alloy 82, it is considered appropriate to apply the combined hardening law.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87989865","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 the present article, authors have carried out a three-dimensional (3D) Computational Fluid Dynamics (CFD) analysis of turbulent natural convection heat transfer from relocated core debris, in typical Sodium cooled Fast Reactors (SFRs), following a Core Disruptive Accident (CDA). Full-Scale analysis of complete sodium pool, i.e., hot and cold pool including immersed decay heat exchanger, has been carried out. k–ω SST model is used for turbulence closure. The model is selected based on the validation exercise. Core catcher (CC) with multiple passive jets over the Heat Shield Plate (HSP) is considered for analysis. Earlier CFD analysis with the assumption of whole core relocation on CC gave a CC temperature higher than the allowable limit. Hence, in this study, the analysis of partial relocation of the core debris on the CC has been carried out. From this, the maximum extent of relocatable core debris on the CC, which conforms with the allowable criteria, has been observed. Therefore, we have investigated the cases where the percentage of core debris relocation varies from 30–100% on HSP and remaining in the original position. This configuration may influence the decay heat removal via the hot pool. Time-dependent decay heat sources are used. Isotherms and streamlines have been presented to understand heat transfer characteristics. It has been found that with the implementation of multi jets CC, debris settled on HSP does not cross the threshold sodium boiling (∼1200 K) temperature up to 70% debris relocated to HSP with single tray configuration. Heat source surface, which remains at the core and in direct contact with coolant (liquid sodium), reaches a maximum value ∼1031 K for the case where the two-third core is intact at the core region. For HSP, it has been found that the thermal design limit exceeds (∼923 K) when 50% of debris relocates to the lower plenum. The transient study shows that time to attain maximum temperature by debris and HSP is inversely proportional to the percentage of intact core.
{"title":"Investigation on Post Accident Heat Removal From Partial Core Relocation in Lower Plenum After CDA in SFRs: 3-D CFD Analysis","authors":"Vidhyasagar Jhade, A. Sharma, N. Kasinathan","doi":"10.1115/icone2020-16682","DOIUrl":"https://doi.org/10.1115/icone2020-16682","url":null,"abstract":"\u0000 In the present article, authors have carried out a three-dimensional (3D) Computational Fluid Dynamics (CFD) analysis of turbulent natural convection heat transfer from relocated core debris, in typical Sodium cooled Fast Reactors (SFRs), following a Core Disruptive Accident (CDA). Full-Scale analysis of complete sodium pool, i.e., hot and cold pool including immersed decay heat exchanger, has been carried out. k–ω SST model is used for turbulence closure. The model is selected based on the validation exercise. Core catcher (CC) with multiple passive jets over the Heat Shield Plate (HSP) is considered for analysis. Earlier CFD analysis with the assumption of whole core relocation on CC gave a CC temperature higher than the allowable limit. Hence, in this study, the analysis of partial relocation of the core debris on the CC has been carried out. From this, the maximum extent of relocatable core debris on the CC, which conforms with the allowable criteria, has been observed. Therefore, we have investigated the cases where the percentage of core debris relocation varies from 30–100% on HSP and remaining in the original position. This configuration may influence the decay heat removal via the hot pool. Time-dependent decay heat sources are used. Isotherms and streamlines have been presented to understand heat transfer characteristics. It has been found that with the implementation of multi jets CC, debris settled on HSP does not cross the threshold sodium boiling (∼1200 K) temperature up to 70% debris relocated to HSP with single tray configuration. Heat source surface, which remains at the core and in direct contact with coolant (liquid sodium), reaches a maximum value ∼1031 K for the case where the two-third core is intact at the core region. For HSP, it has been found that the thermal design limit exceeds (∼923 K) when 50% of debris relocates to the lower plenum. The transient study shows that time to attain maximum temperature by debris and HSP is inversely proportional to the percentage of intact core.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88333448","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}
CANDU PHWR spent fuel pools (SFPs), smaller than a tennis court, contain up to 38,000 or more (49,000 in Wolsong)fuel bundles in geometries not replicated in any other power reactor. Therefore, the phenomenological issues, accident progression pathways and effectiveness of mitigative actions are somewhat different. This requires a dedicated approach in progression and consequence assessments of potential accidents and development of mitigation measures. The SFPs house densely packed fuel bundles stacked in about a hundred vertical stainless steel tray towers, each containing 24 spent fuel bundles in each of the 16 or more (19 in Wolsong) horizontal fish basket style steel trays. Some of theupto 10 year worth of the on-line refuelled bundles in the SFP are at relatively high decay powers as fuel trays are prepped for the towers in near daily basis. In addition, there is a provision (see Figure 1) that a full core of bundles 20 days after being at full power can be transferred to the spent fuel bay at any time. About 4.5m of additional water layer on top of the tray towers provide radiation protection and a healthy margin to small rate of fluid loss.
{"title":"FUELPOOL: A Computer Program to Model CANDU Spent Fuel Pool Severe Accident Progression and Consequences","authors":"Y. Song, Jong-Yeob Jung, S. Nijhawan","doi":"10.1115/icone2020-16634","DOIUrl":"https://doi.org/10.1115/icone2020-16634","url":null,"abstract":"\u0000 CANDU PHWR spent fuel pools (SFPs), smaller than a tennis court, contain up to 38,000 or more (49,000 in Wolsong)fuel bundles in geometries not replicated in any other power reactor. Therefore, the phenomenological issues, accident progression pathways and effectiveness of mitigative actions are somewhat different. This requires a dedicated approach in progression and consequence assessments of potential accidents and development of mitigation measures. The SFPs house densely packed fuel bundles stacked in about a hundred vertical stainless steel tray towers, each containing 24 spent fuel bundles in each of the 16 or more (19 in Wolsong) horizontal fish basket style steel trays. Some of theupto 10 year worth of the on-line refuelled bundles in the SFP are at relatively high decay powers as fuel trays are prepped for the towers in near daily basis. In addition, there is a provision (see Figure 1) that a full core of bundles 20 days after being at full power can be transferred to the spent fuel bay at any time. About 4.5m of additional water layer on top of the tray towers provide radiation protection and a healthy margin to small rate of fluid loss.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79479130","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}
Byung-Hee Choi, D. Orea, Thien Nguyen, N. Anand, Y. Hassan, P. Sabharwall
Texas A&M University is working on the development of gas cooled fast reactor cartridge loop under the Department of Energy VTR program. Our research project aims to develop and implement techniques to quantify the transport and deposition of particle inside the cartridge loop. Before the developed techniques are applied in a complicated actual facility, it is essential to verify and validate their performance using numerical simulations and to quantify their uncertainties. This article presents a numerical study of particle transport and deposition in a proof-of-concept facility. The proof-of-concept facility houses a series of three square duct test sections, each of which has a cross-section of 3 in.2 and a length of 24 in., for a combined total length of 72 in. The numerical simulation domain is based on the geometrical dimensions of the experimental facility. The main stream in the channel is solved using the Eulerian turbulence model, and the particle motion is interpreted in the Lagrangian framework. It is assumed that a well-mixed air–particle mixture at a constant temperature is injected into the horizontal channel. Lagrangian simulations of surrogate particles allow us to understand their behavior precisely. The Reynolds stress model is selected to reproduce the secondary flow and the associated secondary drag force. The state-of-the-art Lagrangian approach, in combination with a random walk model coupled with a computational fluid dynamics model, is employed to investigate the behaviors of the surrogate particles within the square channel. Gravitational settling is also considered. The deposition velocity and penetration efficiency are estimated for representing the characteristics of particle deposition in the proof-of-concept facility. Because the conventional method of measuring the deposition velocity is based on the Eulerian framework, it is not suitable for direct adoption in the Lagrangian framework. This study proposes a numerical technique to measure the deposition velocity; this technique can be efficiently used in the Lagrangian framework of the simulation. The results agree well with both our experimental measurements and correlations available in the literature. Using this technique, the correlations for the deposition velocity are established as functions of the normalized channel length, Stokes number, and Reynolds number. Finally, the relationship between the deposition velocity and penetration efficiency is examined, and a correlation is proposed. Consequently, the penetration efficiency can be directly compared with several conventional reference data based on the deposition velocity.
{"title":"Deposition Velocity and Penetration Efficiency in a Square Channel Using a Lagrangian-Based Modeling Approach","authors":"Byung-Hee Choi, D. Orea, Thien Nguyen, N. Anand, Y. Hassan, P. Sabharwall","doi":"10.1115/icone2020-16907","DOIUrl":"https://doi.org/10.1115/icone2020-16907","url":null,"abstract":"\u0000 Texas A&M University is working on the development of gas cooled fast reactor cartridge loop under the Department of Energy VTR program. Our research project aims to develop and implement techniques to quantify the transport and deposition of particle inside the cartridge loop. Before the developed techniques are applied in a complicated actual facility, it is essential to verify and validate their performance using numerical simulations and to quantify their uncertainties. This article presents a numerical study of particle transport and deposition in a proof-of-concept facility.\u0000 The proof-of-concept facility houses a series of three square duct test sections, each of which has a cross-section of 3 in.2 and a length of 24 in., for a combined total length of 72 in. The numerical simulation domain is based on the geometrical dimensions of the experimental facility. The main stream in the channel is solved using the Eulerian turbulence model, and the particle motion is interpreted in the Lagrangian framework. It is assumed that a well-mixed air–particle mixture at a constant temperature is injected into the horizontal channel. Lagrangian simulations of surrogate particles allow us to understand their behavior precisely.\u0000 The Reynolds stress model is selected to reproduce the secondary flow and the associated secondary drag force. The state-of-the-art Lagrangian approach, in combination with a random walk model coupled with a computational fluid dynamics model, is employed to investigate the behaviors of the surrogate particles within the square channel. Gravitational settling is also considered.\u0000 The deposition velocity and penetration efficiency are estimated for representing the characteristics of particle deposition in the proof-of-concept facility. Because the conventional method of measuring the deposition velocity is based on the Eulerian framework, it is not suitable for direct adoption in the Lagrangian framework. This study proposes a numerical technique to measure the deposition velocity; this technique can be efficiently used in the Lagrangian framework of the simulation. The results agree well with both our experimental measurements and correlations available in the literature. Using this technique, the correlations for the deposition velocity are established as functions of the normalized channel length, Stokes number, and Reynolds number. Finally, the relationship between the deposition velocity and penetration efficiency is examined, and a correlation is proposed. Consequently, the penetration efficiency can be directly compared with several conventional reference data based on the deposition velocity.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82829177","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 Prismatic-core Advanced High Temperature Reactor (PAHTR) is a very high temperature reactor type which is one of promising reactor type technologies classified as Generation IV by the International Forum. The new technology designs are identified as being proliferation resistant, safe, economical, efficient, and long fuel cycle. In this paper, the continuous-energy Monte Carlo method is capable of capturing all of the necessary reactor physics parameters using high fidelity two dimensional model with Serpent Monte Carlo code, and applied for a large scale reactor core loaded with TRi-structural ISOtropic (TRISO) particle by taking into account the double heterogeneity effect. These analyses were performed for PAHTR reactor core that utilizes TRISO particles fuel embedded in graphite matrix by applying a new innovative idea of adding Integral Fuel Burnable Absorber (IFBA) as an additional coating layer with a designated thickness. Adding IFBA coating could lead to compressed excess reactivity at the Beginning of Cycle (BOC), and extended burnup cycle. The additional IFBA coating layer is placed in the outer surface of the fuel kernel and covered by the buffer layers that compose the TRISO fuel particle. Neutronic calculations were performed for both TRISO particle unit cell and for full core with homogenous distribution of IFBA coating.
{"title":"Two-Dimensional Full Core Analysis of IFBA-Coated TRISO Fuel Particles in Very High Temperature Reactors","authors":"M. Alrwashdeh, S. Alameri","doi":"10.1115/icone2020-16838","DOIUrl":"https://doi.org/10.1115/icone2020-16838","url":null,"abstract":"\u0000 The Prismatic-core Advanced High Temperature Reactor (PAHTR) is a very high temperature reactor type which is one of promising reactor type technologies classified as Generation IV by the International Forum. The new technology designs are identified as being proliferation resistant, safe, economical, efficient, and long fuel cycle. In this paper, the continuous-energy Monte Carlo method is capable of capturing all of the necessary reactor physics parameters using high fidelity two dimensional model with Serpent Monte Carlo code, and applied for a large scale reactor core loaded with TRi-structural ISOtropic (TRISO) particle by taking into account the double heterogeneity effect. These analyses were performed for PAHTR reactor core that utilizes TRISO particles fuel embedded in graphite matrix by applying a new innovative idea of adding Integral Fuel Burnable Absorber (IFBA) as an additional coating layer with a designated thickness. Adding IFBA coating could lead to compressed excess reactivity at the Beginning of Cycle (BOC), and extended burnup cycle. The additional IFBA coating layer is placed in the outer surface of the fuel kernel and covered by the buffer layers that compose the TRISO fuel particle. Neutronic calculations were performed for both TRISO particle unit cell and for full core with homogenous distribution of IFBA coating.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75880902","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}
This paper provides the seismic time history data precision and time interval requirement for seismic dynamic analysis. U.S.NRC SRP 3.7.1 “Seismic Design Parameters” Acceptance Criteria for Design Time Histories specifies the power spectral density Nyquist Frequency, time interval, and total duration; however, it does not have the requirement for Response Spectra. The response spectrum bandwidth is inverse-proportional to time interval of the time history. For the time interval of 0.005 seconds, the bandwidth for the response spectrum is between 0.194 Hz and 80.5 Hz; the PSD Nyquist frequency is 100 Hz. For 20.48 seconds time history, 4096 data points are required. The response spectrum between 1.28 Hz and 13.6 Hz has the peak flat magnitude value; the magnitude drops to 0.707 of the peak value from 1.28 Hz to 0.194 Hz and from 13.6 Hz to 80.5 Hz. This paper also provides the time interval requirement for various response spectrum peak flat magnitude value; i.e., the response spectrum highest flat magnitude of 27.2 Hz requires a time interval of 0.0025 seconds time history. For 20.48 seconds time history, 8192 data points are required. For CSDRS, the time interval of 0.005 seconds is adequate for the frequency range of interest between 0.36 Hz and 57.2 Hz. For HRHF, the time interval of 0.0025 seconds is required to analyze the frequency range of interest between 0.36 Hz and 114.4 Hz.
{"title":"Seismic Time History Data Precision and Time Interval Requirement","authors":"Dali Li","doi":"10.1115/icone2020-16594","DOIUrl":"https://doi.org/10.1115/icone2020-16594","url":null,"abstract":"\u0000 This paper provides the seismic time history data precision and time interval requirement for seismic dynamic analysis. U.S.NRC SRP 3.7.1 “Seismic Design Parameters” Acceptance Criteria for Design Time Histories specifies the power spectral density Nyquist Frequency, time interval, and total duration; however, it does not have the requirement for Response Spectra. The response spectrum bandwidth is inverse-proportional to time interval of the time history. For the time interval of 0.005 seconds, the bandwidth for the response spectrum is between 0.194 Hz and 80.5 Hz; the PSD Nyquist frequency is 100 Hz. For 20.48 seconds time history, 4096 data points are required. The response spectrum between 1.28 Hz and 13.6 Hz has the peak flat magnitude value; the magnitude drops to 0.707 of the peak value from 1.28 Hz to 0.194 Hz and from 13.6 Hz to 80.5 Hz. This paper also provides the time interval requirement for various response spectrum peak flat magnitude value; i.e., the response spectrum highest flat magnitude of 27.2 Hz requires a time interval of 0.0025 seconds time history. For 20.48 seconds time history, 8192 data points are required. For CSDRS, the time interval of 0.005 seconds is adequate for the frequency range of interest between 0.36 Hz and 57.2 Hz. For HRHF, the time interval of 0.0025 seconds is required to analyze the frequency range of interest between 0.36 Hz and 114.4 Hz.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82039145","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}
M. Sonehara, Mitsuhiro Aoyagi, A. Uchibori, T. Takata, H. Ohshima, A. Clark, D. Louie
In order to investigate the effect of sodium combustion, Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) have exchanged information of sodium combustion modeling and related experimental data in the framework of Civil Nuclear Energy Research and Development Working Group (CNWG). This collaboration includes a benchmark analysis of the SNL Surtsey spray combustion experiment (SNL T3 experiments) using AQUA-SF and SPHINCS in JAEA. In this paper, investigation into multi-dimensional effect and best estimate for T3 experiment with AQUA-SF are conducted as validation and verification of the code. A spray combustion is characterized by formation of sodium droplet cloud due to pressure difference and their spreading with combustion. Therefore, the combustion phenomenon will be much affected by spatial distributions of parameters such as gas temperature, gas velocity and oxygen concentration. As a best estimate analysis, the spray burning duration is adjusted in the computation in order to take into account the temporary suppression of the spray combustion observed in the experiment. Furthermore, droplet size of SPHINCS and AQUA-SF are optimized to represent the T3 experimental results. For the best estimate in AQUA-SF, sodium droplet size needs to be set larger than SPHINCS in order to decrease the surface area and suppress the spray burning rate. These adjustment leads to more precise representation of the measurements in T3 experiment.
{"title":"Numerical Validation of AQUA-SF in SNL T3 Sodium Spray Fire Experiment","authors":"M. Sonehara, Mitsuhiro Aoyagi, A. Uchibori, T. Takata, H. Ohshima, A. Clark, D. Louie","doi":"10.1115/icone2020-16851","DOIUrl":"https://doi.org/10.1115/icone2020-16851","url":null,"abstract":"\u0000 In order to investigate the effect of sodium combustion, Sandia National Laboratories (SNL) and Japan Atomic Energy Agency (JAEA) have exchanged information of sodium combustion modeling and related experimental data in the framework of Civil Nuclear Energy Research and Development Working Group (CNWG). This collaboration includes a benchmark analysis of the SNL Surtsey spray combustion experiment (SNL T3 experiments) using AQUA-SF and SPHINCS in JAEA. In this paper, investigation into multi-dimensional effect and best estimate for T3 experiment with AQUA-SF are conducted as validation and verification of the code. A spray combustion is characterized by formation of sodium droplet cloud due to pressure difference and their spreading with combustion. Therefore, the combustion phenomenon will be much affected by spatial distributions of parameters such as gas temperature, gas velocity and oxygen concentration.\u0000 As a best estimate analysis, the spray burning duration is adjusted in the computation in order to take into account the temporary suppression of the spray combustion observed in the experiment. Furthermore, droplet size of SPHINCS and AQUA-SF are optimized to represent the T3 experimental results. For the best estimate in AQUA-SF, sodium droplet size needs to be set larger than SPHINCS in order to decrease the surface area and suppress the spray burning rate. These adjustment leads to more precise representation of the measurements in T3 experiment.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80944736","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}
R. Marinari, I. Piazza, M. Tarantino, G. Grasso, M. Frignani
In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the coolability of the Fuel Assembly in nominal condition is of central interest. The Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) is a 300 MWth pool-type reactor aimed at demonstrating the safe deployment of the Generation IV LFR technology. The ALFRED design, currently being developed by the Fostering ALFRED Construction international consortium, is based on prototypical solutions intended to be used in the next generation of lead-cooled Small Modular Reactors. Within the scope of FALCON and in the frame of investigating the thermal-hydraulics of the ALFRED core, a CFD computational model of the general Fuel Assembly (FA) is built looking for the assessment of its thermal field in nominal flow conditions both for the average FA and the hottest one. Starting from the experience in this kind of simulations and in experimental work, the whole model of the ALFRED Fuel Assembly is first presented and calculation of flow and temperature field in nominal conditions is carried out. Results showed that the thermal hydraulic field predicted in the average FA by the code is in good agreement with analytical correlations and the temperature field on the pin clad is acceptable for clad material temperature constraint. About the results on the hot FA test case, the CFD results highlighted a peak temperature on the clad close to the clad temperature constraint. This result led to an upgrade of the mass flow distribution among the FA for achieving a 20% mass flow increase in the hottest one that guarantees higher temperature margin on the clad.
{"title":"CFD Preliminary Assessment of the ALFRED FA Thermal-Hydraulics","authors":"R. Marinari, I. Piazza, M. Tarantino, G. Grasso, M. Frignani","doi":"10.1115/icone2020-16593","DOIUrl":"https://doi.org/10.1115/icone2020-16593","url":null,"abstract":"\u0000 In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the coolability of the Fuel Assembly in nominal condition is of central interest.\u0000 The Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED) is a 300 MWth pool-type reactor aimed at demonstrating the safe deployment of the Generation IV LFR technology. The ALFRED design, currently being developed by the Fostering ALFRED Construction international consortium, is based on prototypical solutions intended to be used in the next generation of lead-cooled Small Modular Reactors.\u0000 Within the scope of FALCON and in the frame of investigating the thermal-hydraulics of the ALFRED core, a CFD computational model of the general Fuel Assembly (FA) is built looking for the assessment of its thermal field in nominal flow conditions both for the average FA and the hottest one. Starting from the experience in this kind of simulations and in experimental work, the whole model of the ALFRED Fuel Assembly is first presented and calculation of flow and temperature field in nominal conditions is carried out. Results showed that the thermal hydraulic field predicted in the average FA by the code is in good agreement with analytical correlations and the temperature field on the pin clad is acceptable for clad material temperature constraint. About the results on the hot FA test case, the CFD results highlighted a peak temperature on the clad close to the clad temperature constraint. This result led to an upgrade of the mass flow distribution among the FA for achieving a 20% mass flow increase in the hottest one that guarantees higher temperature margin on the clad.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74989941","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}
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