scholarly journals Thermal-hydraulic modeling of reactivity accidents in MTR reactors

2006 ◽  
Vol 21 (2) ◽  
pp. 21-32 ◽  
Author(s):  
Hany Khater ◽  
Talal Abu-El-Maty ◽  
El-Din El-Morshdy
Keyword(s):  
Present Model ◽  
Reactor Core ◽  
Hydraulic Modeling ◽  
Control Rod ◽  
Reactor Kinetics ◽  
Fortran Language ◽  
Thermal Hydraulic Modeling ◽  
Thermal Hydraulic Analysis ◽  
Reactivity Insertion ◽  
Good Agreement

This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.

Thermal Hydraulic Analysis of Severe Accident in PFBR

2010 ◽  
Author(s):  
K. Velusamy ◽  
P. Chellapandi ◽  
G. R. Raviprasan ◽  
P. Selvaraj ◽  
S. C. Chetal
Keyword(s):  
Atmospheric Pressure ◽  
Fluid Dynamic ◽  
Hydraulic Modeling ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Maximum Pressure ◽  
Dynamic Calculation ◽  
Thermal Hydraulic Modeling ◽  
Thermal Hydraulic Analysis ◽  
Design Pressure

During a core disruptive accident (CDA), the amount of primary sodium that can be released to Reactor Containment Building (RCB) in Prototype Fast Breeder Reactor (PFBR) is estimated to be 350 kg/s, by a transient fluid dynamic calculation. The pressure and temperature evolutions inside RCB, due to consequent sodium fire have been estimated by a constant burning rate model, accounting for heat absorption by RCB wall, assuming RCB isolation based on area gamma monitors. The maximum pressure developed is 7000 Pa. In case RCB isolation is delayed, then the final pressure inside RCB reduces below atmospheric pressure due to cooling of RCB air. The negative pressure that can be developed is estimated by dynamic thermal hydraulic modeling of RCB air / wall to be −3500 Pa. These investigations were useful to arrive at the RCB design pressure. Following CDA, RCB is isolated for 40 days. During this period, the heat added to RCB is dissipated to atmosphere only by natural convection. Considering all the possible routes of heat addition to RCB, evolution of RCB wall temperature has been predicted using HEATING5 code. It is established that the maximum temperature in RCB wall is less than the permissible value.

scholarly journals ANALYSIS OF THE SUPERPHENIX START-UP TESTS WITH APOLLO-3: FROM ZERO POWER ISOTHERMAL CONDITIONS TO DYNAMIC POWER TRANSIENT ANALYSIS

2021 ◽  
Vol 247 ◽  
pp. 06044
Author(s):  
Sciora Pierre ◽  
Garcia Elias ◽  
Rimpault Gerald ◽  
Droin Jean-Baptiste ◽  
Pascal Vincent
Keyword(s):  
Potential Candidate ◽  
Neutron Physics ◽  
Control Rod ◽  
Isothermal Conditions ◽  
Energy Needs ◽  
Start Up ◽  
Experimental Facilities ◽  
Irradiation Behavior ◽  
Reactivity Insertion ◽  
Good Agreement

Sodium-cooled Fast Reactors (SFRs) remain a potential candidate to meet future energy needs. In addition, the SFRs experimental feedback is considerable, for instance, the French research program has considered experimental facilities including the Superphénix which has emerged as a transition to commercial deployment. In this paper a set of tests from the Superphénix start-up are reanalyzed with new tools, considering APOLLO-3 and TRIPOLI-4 (respectively deterministic and stochastic codes) for neutron physics evaluation, GERMINAL-V2 for the fuel irradiation behavior and CATHARE-3 for the thermal-hydraulics modelling. Neutron physics evaluations are performed for the main control rod worth and the Doppler Effect, both measured under isothermal conditions at Superphénix start-up. A good agreement is obtained for these tests, which were purely neutronic tests. Next, the core temperature distribution is evaluated at nominal conditions, where larger discrepancies are observed. However, these deviations are related to the measurement of the fuel assemblies, which have a larger than expected uncertainty. Finally a transient, consisting of a negative reactivity insertion, is analyzed to assess the dynamic core behavior. A good agreement is obtained during the reactivity insertion, however the thermal-hydraulic model has to be improved, namely the vessel model, which is considered as a 0-D volume.

scholarly journals ANALYSIS OF REACTIVITY INSERTION AS A FUNCTION OF THE RSG-GAS FUEL BURN-UP

2021 ◽  
Vol 23 (1) ◽  
pp. 1
Author(s):  
Tukiran Surbakti ◽  
Surian Pinem ◽  
Lily Suparlina
Keyword(s):  
Reactor Core ◽  
Static Condition ◽  
Power Coefficient ◽  
Maximum Temperature ◽  
Control Rod ◽  
Chain Reactions ◽  
The Core ◽  
Fuel Burn ◽  
Reactivity Insertion ◽  
Safety Limit

Analysis of the control rod insertion is important as it is closely related to reactor safety. Previously, the analysis has been carried out in RSG-GAS during static condition, not as a function of the fuel fraction. The RSG-GAS reactor in one cycle is a function of the fuel burn-up. It is necessary to analyze RSG-GAS core reactivity insertion as a function of the fuel burn-up to determine the behavior of the reactor, especially in uncontrolled operations such as continuous pulling of control rods. This analysis is carried out by the computer simulation method using WIMSD-5B and MTR-DYN codes, by observing power behavior as a function of time due to neutron chain reactions in the reactor core. Calculations are performed using point kinetics equation, and the feedback effect will be evaluated using static power coefficient and fuel burn-up function. Analyzes were performed for the core configuration of the core no. 99, by lifting the control rod or inserting positive reactivity to the core. The calculation results show that with the reactivity insertion of 0.5% Δk/k at start-up power of 1 W and 1 MW, safety limit is not exceeded either at the beginning, middle, or end of the cycle. The maximum temperature of the fuel is 135°C while the safety limit is 180°C. The margin from the safety limit is large, and therefore fuel damage is not possible when power excursion were to occur.

Comparison Between PARCS and MCNP6 Codes on VVER1000/V320 Core

2014 ◽  
Author(s):  
Guido Mazzini ◽  
Bruno Miglierini ◽  
Marek Ruščák
Keyword(s):  
Cross Sections ◽  
Three Dimensional ◽  
Reactor Core ◽  
Research Centre ◽  
Control Rod ◽  
Neutron Kinetics ◽  
Reactor Kinetics ◽  
Design Basis ◽  
Accident Scenarios ◽  
Fully Functioning

Research Centre Rez solves several safety related projects dealing with safety of Czech NPPs, some of which require fully functioning Three Dimensional (3D) model of the reactor core. While in a number of safety analysis of various accident scenarios it is sufficient to use one point reactor kinetics, there are selected types of accidents in which it is useful to model the space (3D) neutron kinetics, in particular control rod ejections, boron dilution scenarios, including transitions from design basis to beyond design basis accidents. This paper is focused to analyze the present model of the core of VVER1000/V320 reactor. Which is applicable for 3D modeling of neutron kinetics in selected design and beyond design basis accidents. The model is based on a cross-sections library created by SCALE 6.1.2/TRITON simulations. PARCS 3.2 code uses homogenized cross-sections libraries to calculate neutronic and other core parameters of the PWR reactors. Similar model is prepared with MCNP6 for comparison between deterministic (Pn spherical-harmonics method used in PARCS) and the stochastic (Monte Carlo) approach (used in MCNP6). Such comparison will serve as a demonstration of the capability of the PARCS code for VVER1000/V320 analyses.

A Compact Model for Spherical Rough Contacts

2004 ◽  
Author(s):  
M. Bahrami ◽  
M. M. Yovanovich ◽  
J. R. Culham
Keyword(s):  
Pressure Distribution ◽  
Entire Range ◽  
Present Model ◽  
Numerical Solutions ◽  
Contact Radius ◽  
Bulk Deformation ◽  
Dimensional Parameters ◽  
Maximum Contact ◽  
Good Agreement ◽  
Key Parameter

The contact of rough spheres is of high interest in many tribological, thermal, and electrical fundamental analyses. Implementing the existing models is complex and requires iterative numerical solutions. In this paper a new model is presented and a general pressure distribution is proposed that encompasses the entire range of spherical rough contacts including the Hertzian limit. It is shown that the non-dimensional maximum contact pressure is the key parameter that controls the solution. Compact expressions are proposed for calculating the pressure distribution, radius of the contact area, elastic bulk deformation, and the compliance as functions of the governing non-dimensional parameters. The present model shows the same trends as those of the Greenwood and Tripp model. Correlations proposed for the contact radius and the compliance are compared with experimental data collected by others and good agreement is observed.

Thermal–hydraulic modeling and transient analysis of helium-cooled tungsten target coupled with ADS core

2016 ◽  
Vol 87 ◽  
pp. 612-620
Author(s):  
Tianji Peng ◽  
Zhiwei Zhou ◽  
Sicong Xiao ◽  
Xuanyu Sheng ◽  
Long Gu
Keyword(s):  
Transient Analysis ◽  
Hydraulic Modeling ◽  
Tungsten Target ◽  
Thermal Hydraulic Modeling

scholarly journals Research on Influence of Different Simulation Methods of Bypass Flow in Thermal Hydraulic Analysis on Temperature Distribution in HTR-10

2020 ◽  
Vol 2020 ◽  
pp. 1-8
Author(s):  
Shiyan Sun ◽  
Youjie Zhang ◽  
Yanhua Zheng
Keyword(s):  
Temperature Distribution ◽  
Reactor Core ◽  
Flow Distribution ◽  
Maximum Temperature ◽  
Flow Ratio ◽  
Hydraulic Analysis ◽  
Bypass Flow ◽  
Pebble Bed ◽  
The Core ◽  
Thermal Hydraulic Analysis

In pebble-bed high temperature gas-cooled reactor, gaps widely exist between graphite blocks and carbon bricks in the reactor core vessel. The bypass helium flowing through the gaps affects the flow distribution of the core and weakens the effective cooling of the core by helium, which in turn affects the temperature distribution and the safety features of the reactor. In this paper, the thermal hydraulic analysis models of HTR-10 with bypass flow channels simulated at different positions are designed based on the flow distribution scheme of the original core models and combined with the actual position of the core bypass flow. The results show that the bypass coolant flowing through the reflectors enhances the heat transfer of the nearby components efficiently. The temperature of the side reflectors and the carbon bricks is much lower with more side bypass coolant. The temperature distribution of the central region in the pebble bed is affected by the bypass flow positions slightly, while that of the peripheral area is affected significantly. The maximum temperature of the helium, the surface, and center of the fuel elements rises as the bypass flow ratio becomes larger, while the temperature difference between them almost keeps constant. When the flow ratio of each part keeps constant, the maximum temperature almost does not change with different bypass flow positions.

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