scholarly journals THERMAL-HYDRAULIC ANALYSIS OF IRT-4M IN REACTIVITY INSERTION ACCIDENT AT VR-1 REACTOR

2016 ◽  
Vol 4 ◽  
pp. 22
Author(s):  
Filip Fejt
Keyword(s):  
Atmospheric Pressure ◽  
Room Temperature ◽  
Nuclear System ◽  
Hydraulic Analysis ◽  
Research Reactors ◽  
The Core ◽  
Fuel Assemblies ◽  
Coolant Circulation ◽  
Thermal Hydraulic Analysis ◽  
Reactivity Insertion

The paper deals with thermal-hydraulic analysis during reactivity insertion accident, i.e. a step increase of nuclear system reactivity by 0.7 eff, at VR-1 Reactor. The reactor utilizes IRT-4M type of fuel assemblies, and even though these fuel assemblies are designed for an operation at the high-power research reactors, they might be also used for zero-power reactors. The thermal-hydraulic analyses must take into account several specific assumptions that are derived from VR-1 reactor specifications. The reactor does not require a forced water flow for a fuel cooling, the core is placed in an open vessel with atmospheric pressure, and amount of coolant water in the vessel is sufficient for providing the inlet water at room temperature for the whole event. Coolant circulation is expected to be formed only by natural convection.

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.

Computational Analysis of Downcomer Boiling Phenomena Using a Component Thermal Hydraulic Analysis Code CUPID

2010 ◽  
Author(s):  
H. K. Cho ◽  
B. J. Yun ◽  
I. K. Park ◽  
J. J. Jeong
Keyword(s):  
Nuclear Reactor ◽  
Reactor Vessel ◽  
Physical Models ◽  
Injection System ◽  
Hydraulic Analysis ◽  
Two Phase ◽  
Rate Condition ◽  
The Core ◽  
Benchmark Calculation ◽  
Thermal Hydraulic Analysis

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analyses of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopts three-dimensional, transient, two-phase and three-field model, and includes various physical models and correlations of the interfacial mass, momentum and energy transfer for the closure relations of the two-fluid model. In the present paper, the two-phase models were assessed against the DOBO (DOwncomer BOiling) experiment, which was constructed to simulate the downcomer boiling phenomenon. It may happen in the downcomer of a nuclear reactor vessel during the reflood phase of a postulated loss of coolant accident. The stored energy release from the reactor vessel to the liquid inside the downcomer causes the boiling on the wall, and it can reduce the hydraulic head of the accumulated water, which is the driving force of water reflooding to the core. This phenomenon has been considered as a crucial safety issue of an advanced power reactor because it is concerned with the core cooling capability of the safety injection system. In this paper, the physical models and correlations that were incorporated into the CUPID code were introduced and the validation results against the experiment were reported. The benchmark calculation results concluded that the CUPID code can appropriately predict the boiling phenomena under a low pressure and low flow rate condition with modification of the bubble size correlation.

scholarly journals Thermal-hydraulic analysis of a VVER-1000 core in MSLB conditions

2021 ◽  
Vol 327 ◽  
pp. 01013
Author(s):  
Svetlomir Mitkov ◽  
Ivan Spasov ◽  
Nikola Kolev
Keyword(s):  
Hydraulic Analysis ◽  
The Core ◽  
Fine Mesh ◽  
Core Boundary ◽  
Safety Parameters ◽  
Equipment Failures ◽  
Main Steam Line ◽  
Main Steam ◽  
Control Rods ◽  
Thermal Hydraulic Analysis

The objective of this paper is to analyze the ability of a VVER-1000 core and its control system to cope with a hypothetical main steam line break (MSLB) accident in case of multiple equipment failures. The study involves the use of advanced 3D core calculation models benchmarked and validated for reactivity accidents in preceding studies. A MSLB core boundary condition problem is solved on a coarse (nodal) mesh with the coupled COBAYA/CTF neutronic/thermal hydraulic codes. The core thermal-hydraulic boundary conditions are obtained from a preceding full-plant MSLB simulation. The assessment of the core safety parameters is supplemented by a fine-mesh (sub-channel) thermal-hydraulic analysis of the hottest assemblies with the CTF code using information from the 3D nodal COBAYA/CTF calculations. Thirteen variants of a pessimistic MSLB scenario are considered, each of them assuming a number of equipment failures aggravated by eight control rods stuck out of the core after scram at different locations in the overcooled sector. The results (within the limitations of the adopted modeling assumptions) show that the core safety parameters do not exceed the safety limits in the simulated aggravated reactivity accidents.

SEISMIC VELOCITY STUDY OF SYNTHETIC CORES

Geophysics ◽  
1957 ◽  
Vol 22 (4) ◽  
pp. 813-820 ◽  
Author(s):  
William O. Murphy ◽  
Joseph W. Berg ◽  
Kenneth L. Cook
Keyword(s):  
Elastic Wave ◽  
Elastic Waves ◽  
Atmospheric Pressure ◽  
Room Temperature ◽  
Seismic Velocity ◽  
Pore Filling ◽  
The Core ◽  
Wave Velocities ◽  
Longitudinal Elastic Wave ◽  
Alcohol Benzene

The velocity of a longitudinal elastic wave through rock at room temperature and at atmospheric pressure depends upon the nature of the rock frame, the porosity of the rock, and the nature of the pore‐filling fluid. In the present investigation longitudinal elastic wave velocities were measured for sixty synthetic cores. The rock frame consisted of sorted quartz sand grains and cement, the percentage of cement varying from ten to fifty percent. The core porosities varied from 8.8 percent to 22 percent. The velocities were measured for dry air‐filled cores and for cores saturated with various liquids. These pore‐filling liquids were distilled water, ethyl alcohol, benzene, carbon tetrachloride, and chloroform. The measured velocities ranged from 2,360 feet per second to 14,710 feet per second. The wave velocity in liquid‐filled cores of 10 percent porosity was approximately twice the velocity for cores of 20 percent porosity, the same type of cement being used in both instances. For any given core, flooded with fluids of wave velocities ranging from 3,000 to 5,000 feet per second, the maximum observed variation in core velocity was around 20 percent. The experimental data fitted the empirical linear equation [Formula: see text] where [Formula: see text] of longitudinal elastic waves passing through the flooded core; [Formula: see text] of longitudinal elastic waves in passing through the saturating fluid. The constant k depends upon the porosity of the rock and the type of cement used. The constant, C, depends upon the nature of the rock frame.

Thermal hydraulic analysis of reactivity accidents in MTR research reactors using RELAP5

Kerntechnik ◽  
2015 ◽  
Vol 80 (6) ◽  
pp. 557-562 ◽  
Author(s):  
N. El-Sahlamy ◽  
A. Khedr ◽  
F. D'Auria
Keyword(s):  
Hydraulic Analysis ◽  
Research Reactors ◽  
Thermal Hydraulic Analysis

scholarly journals Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

2011 ◽  
Vol 26 (1) ◽  
pp. 45-49 ◽  
Author(s):  
Atta Muhammad ◽  
Masood Iqbal ◽  
Tayyab Mahmood
Keyword(s):  
Steady State ◽  
Research Reactor ◽  
Safety Margin ◽  
Nucleate Boiling ◽  
Hydraulic Analysis ◽  
The Core ◽  
Computer Codes ◽  
Neutronic Parameters ◽  
Thermal Hydraulic Analysis

The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

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