Effect of Ocean Conditions on Neutronic/Thermal-Hydraulic Coupling of IPWR

2018 ◽  
Author(s):  
Genglei Xia
Keyword(s):  
Natural Circulation ◽  
Oscillation Amplitude ◽  
Three Dimensional ◽  
Simulation Models ◽  
Reactor Power ◽  
Coolant Flow ◽  
Uneven Distribution ◽  
Thermal Hydraulics ◽  
Primary Coolant ◽  
Ocean Conditions

In this paper, a thermal-hydraulics analysis code under non-inertial system was developed by establishing dynamic simulation models of typical ocean conditions and modifying RELAP5 code. Based on the modified code, the effects of the inclination, fluctuation and rolling conditions on the natural circulation characteristics of an integrated PWR were studied. The dynamic input and output interface model of RELAP5 was further established, and RELAP5 was coupled with a two-group three-dimensional neutron kinetics code to realize the improvement of nuclear feedback module in thermal hydraulic code. And the distribution of core flow and power under different motions was analyzed. The results indicate that, the uneven distribution of coolant flow increases with the increasing inclination angle, this trend leads to an uneven distribution of primary coolant flow. The flow fluctuation has a 180° phase difference under rolling conditions, and the reactor power and coolant flow oscillation increases with the increasing rolling period and amplitude. In the case of heaving motion, the peaks of the oscillation amplitude of the flow and power lying in the hottest channel as the additional forces on the fluid of each channel are spatially uniform. The code developed in this paper has the functions of modelling ocean conditions and three-dimensional coupled neutronics/thermal-hydraulics, and can be used as a simulation tool for Floating PWR.

Analysis for the Case of Channel Flow and Reactivity Changes for Advanced Heavy Water Reactor

2004 ◽  
Author(s):  
A. Srivastava ◽  
P. Majumdar ◽  
D. Mukhopadhyay ◽  
H. G. Lele ◽  
S. K. Gupta
Keyword(s):  
Heavy Water ◽  
Natural Circulation ◽  
Heat Removal ◽  
Coolant Flow ◽  
Thermal Hydraulics ◽  
Primary Coolant ◽  
Control Rod ◽  
Water Reactor ◽  
Wide Range ◽  
Heavy Water Reactor

The proposed Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power level with no primary coolant pumps. Decrease in coolant flow or control rod malfunction can lead to undesirable rise in clad surface temperature depending upon severity and characteristics and response of the reactor and associated systems. In this paper safety assessment of the AHWR is made due to above events of different severity. Cause for events under category of decrease in coolant flow is mainly channel blockage of different severity at different locations. There is no other reason as it is natural circulation based reactor. Effect of flow decrease can be different in different channels and at different axial locations. In this paper channel blockages of different sizes are analysed at core inlet and using slave channel approach. Changes in reactivities can occur due to inadvertent withdrawal of one or more control rods from reactor core. In this analysis one control rod assembly is assumed to be removed from core. The event is simulated by addition of 5 mk reactivity in 120 seconds depending on the speed of withdrawal of assembly. The analysis for the above events are complex due to various complex and wide range of phenomena involved during different PIEs under this category. It involves single and two phase natural circulation at different power levels, inventories and pressures, coupled neutronics and thermal hydraulics behaviour, and coupled controller and thermal hydraulics. In this paper summary of analysis for each event is presented. In this paper, various modeling complexities are brought out; evaluation of acceptance criteria is made and design implications of each event are discussed.

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

2006 ◽  
Author(s):  
A. Gorzel
Keyword(s):  
Three Dimensional ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressure Increase ◽  
Thermal Hydraulics ◽  
Short Term ◽  
Speed Reduction ◽  
Pump Speed ◽  
Transient Behaviour ◽  
Limit Value

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and impermissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second — much smaller — maximum that would occur around one second after the first one in the absence of a SCRAM.

On the MEGAPIE Target Thermal Hydraulics: A RELAP5 Analysis

2004 ◽  
Author(s):  
W. H. Leung
Keyword(s):  
Proton Beam ◽  
Natural Circulation ◽  
Heat Removal ◽  
Target Material ◽  
Thermal Hydraulics ◽  
Primary Coolant ◽  
Removal Capacity ◽  
Circulation Mode ◽  
Long Run ◽  
Cooling Loops

A new type of liquid-metal target is designed for the Spallation Neutron Source in PSI. LBE is selected to be the target material and the primary coolant as well. RELAP5/MOD 3.2 is used to analyze the system thermal hydraulics. The nominal conditions are chosen based on temperature constraints from the design assessments. The steady state results are in the proximity of the design specifications and the heat removal capacity is adequately deployed. The normal thermal hydraulic transients, namely the proton beam and beam interrupt, are studied. A basic PID (Proportional, Integral, Derivative) control is implemented in the RELAP5 for regulating the target temperature. It is found that the control chain works very well for the beam trip in limiting the temperature fluctuations. In a beam interrupt, the proton beam is completely turned off without recovering. The transition from full power to hot-standby is quite smooth, but it becomes oscillatory in the long run due to the timelags in the cooling loops’ responses. An off normal case of target main coolant trip has also been studied. Without the main pump, the target can still be operated in the natural circulation mode, and the control can cope with the normal beam transients and restarting the target from hot standby.

Effect of Outer Fin Axial Gap on the Sealing Effectiveness and Fluid Dynamics of Radial Rim Seal

2017 ◽  
Author(s):  
Zhigang Li ◽  
Jun Li ◽  
Liming Song ◽  
Qing Gao ◽  
Xin Yan ◽  
...  
Keyword(s):  
Fluid Dynamics ◽  
Flow Rate ◽  
Gas Turbines ◽  
Flow Behavior ◽  
Three Dimensional ◽  
Numerical Approach ◽  
Coolant Flow ◽  
Coolant Flow Rate ◽  
Navier Stokes ◽  
Hot Gas

The modern gas turbine is widely applied in the aviation propulsion and power generation. The rim seal is usually designed at the periphery of the wheel-space and prevented the hot gas ingestion in modern gas turbines. The high sealing effectiveness of rim seal can improve the aerodynamic performance of gas turbines and avoid of the disc overheating. Effect of outer fin axial gap of radial rim seal on the sealing effectiveness and fluid dynamics was numerically investigated in this work. The sealing effectiveness and fluid dynamics of radial rim seal with three different outer fin axial gaps was conducted at different coolant flow rates using three-dimensional Reynolds-Averaged Navier-Stokes (RANS) and SST turbulent model solutions. The accuracy of the presented numerical approach for the prediction of the sealing performance of the turbine rim seal was demonstrated. The obtained results show that the sealing effectiveness of radial rim seal increases with increase of coolant flow rate at the fixed axial outer fin gap. The sealing effectiveness increases with decrease of the axial outer fin gap at the fixed coolant flow rate. Furthermore, at the fixed coolant flow rate, the hot gas ingestion increases with the increase of the axial outer fin gap. This flow behavior intensifies the interaction between the hot gas and coolant flow at the clearance of radial rim seal. The preswirl coefficient in the wheel-space cavity is also illustrated to analyze the flow dynamics of radial rim seal at different axial outer fin gaps.

scholarly journals The Design of Turbomachinery for the Gas Turbine (Direct Cycle) High Temperature Gas Cooled Reactor Power Plant

1977 ◽  
Author(s):  
R. G. Adams ◽  
F. H. Boenig
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Gas Turbine ◽  
Power Conversion ◽  
Reactor Power ◽  
Working Fluid ◽  
Primary Coolant ◽  
Reactor Power Plant ◽  
Direct Cycle ◽  
High Temperature Gas

The Gas Turbine HTGR, or “Direct Cycle” High-Temperature Gas-Cooled, Reactor power plant, uses a closed-cycle gas turbine directly in the primary coolant circuit of a helium-cooled high-temperature nuclear reactor. Previous papers have described configuration studies leading to the selection of reactor and power conversion loop layout, and the considerations affecting the design of the components of the power conversion loop. This paper discusses briefly the effects of the helium working fluid and the reactor cooling loop environment on the design requirements of the direct-cycle turbomachinery and describes the mechanical arrangement of a typical turbomachine for this application. The aerodynamic design is outlined, and the mechanical design is described in some detail, with particular emphasis on the bearings and seals for the turbomachine.

A Model of Complexity Levels of Meaning Constitution in Simulation Models of Language Evolution

2011 ◽  
Vol 1 (1) ◽  
pp. 18-38 ◽  
Author(s):  
Andy Lücking ◽  
Alexander Mehler
Keyword(s):  
Physical Properties ◽  
Language Evolution ◽  
Three Dimensional ◽  
Simulation Models ◽  
Context Dependency ◽  
Abstract Objects ◽  
Linguistic Terms ◽  
Left And Right ◽  
Linguistic Categories ◽  
Linguistic Categorization

Currently, some simulative accounts exist within dynamic or evolutionary frameworks that are concerned with the development of linguistic categories within a population of language users. Although these studies mostly emphasize that their models are abstract, the paradigm categorization domain is preferably that of colors. In this paper, the authors argue that color adjectives are special predicates in both linguistic and metaphysical terms: semantically, they are intersective predicates, metaphysically, color properties can be empirically reduced onto purely physical properties. The restriction of categorization simulations to the color paradigm systematically leads to ignoring two ubiquitous features of natural language predicates, namely relativity and context-dependency. Therefore, the models for simulation models of linguistic categories are not able to capture the formation of categories like perspective-dependent predicates ‘left’ and ‘right’, subsective predicates like ‘small’ and ‘big’, or predicates that make reference to abstract objects like ‘I prefer this kind of situation’. The authors develop a three-dimensional grid of ascending complexity that is partitioned according to the semiotic triangle. They also develop a conceptual model in the form of a decision grid by means of which the complexity level of simulation models of linguistic categorization can be assessed in linguistic terms.

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Wang Lianjie ◽  
Lu Di ◽  
Zhao Wenbo
Keyword(s):  
Electric Power ◽  
Supercritical Water ◽  
Transient Analysis ◽  
Three Dimensional ◽  
Thermal Hydraulics ◽  
Water Density ◽  
Fuel Temperature ◽  
Control Rod ◽  
Reactor Protection System ◽  
Supercritical Water Cooled Reactor

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.

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