scholarly journals RESEARCH THE BEHAVIOUR AND PROPERTIES OF WWER TYPE FUEL CLADDINGS FROM Zr1%Nb ALLOY IN LOSS OF THE COOLANT ACCIDENT

2021 ◽  
pp. 80-86
Author(s):  
M.M. Semerak ◽  
S.S. Lys
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Experimental Studies ◽  
Primary Circuit ◽  
Core Flooding ◽  
Acceptance Criteria ◽  
Fuel Rod ◽  
The Core ◽  
Loss Of Coolant ◽  
Type Fuel

The paper presents safety criteria placed on fuel rod condition in loss of the coolant accident (LOCA) conditions as applied to reactor plants with WWER. The paper reveals briefly experimental studies carried out to validate safety criteria (acceptance criteria). The scope of the data experimentally obtained by research the behaviour and properties of WWER type fuel claddings from Zr1%Nb alloy under loading conditions simulating the stage of core flooding with water in LOCA suffices to judge the character and numerical value of criterional parameters of the embrittlement criterion in terms of the cladding stability upon flooding and subsequent implementation of fuel assembly (FA) unloading and transportation. Accidents are considered involving loss of coolant by primary circuit which are characterized by conditions of degraded heat transfer from fuels. During accidents loss of tightness by fuel rod cladding is tolerable, however, in this case, the cooling of a distorted fuel rod and the dismantling (unloading) of the core after an accident have to be feasible.

Numerical Analysis of the Flow and Heat Transfer in the Sub-Channel of Supercritical Water Reactor

2013 ◽  
Author(s):  
Ajoy Debbarma ◽  
K. M. Pandey
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Supercritical Water ◽  
High Performance ◽  
Flow Distribution ◽  
Diameter Ratio ◽  
Light Water Reactor ◽  
Water Reactor ◽  
Fuel Rod ◽  
Gap Width

Research activities are ongoing for High performance light water reactor (HPLWR) with square double rows fuel assembly to develop nuclear power plants with the purpose to achieve a high thermal efficiency and to improve their economical competitiveness. However, there is still a big deficiency in understanding and prediction of heat transfer in supercritical fluids. This paper evaluates three-dimensional turbulent flow and convective heat transfer in a single-phase and steady-state sub-channel of HPLWR by using general computational fluid dynamics code, Ansys 14 Fluent. The major concern using supercritical water as work fluid is the heat transfer characteristics due to large variations of thermal properties of supercritical water near pseudo-critical line. In order to ensure the safety of operation in High performance light water reactor (HPLWR), heat transfer deterioration (HTD) must be avoided. Numerical results prove that the RNG k-e model with the enhanced near-wall treatment obtained the most satisfactory prediction and lead to satisfactory simulation results. The HPLWR Square fuel assembly has many square-shaped water rods, Out of four types of sub-channels; three sub-channels SC-1, SC-2 and SC-3 are investigated (adjacent to the side of the moderator flow channels (SC-1) (moderator tube and assembly gap), central sub-channels formed by four fuel rods (SC-2), adjacent to the corner of the moderator tube (SC-3). Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and water rod, fuel rod pitch to diameter ratio 1.1–1.4 with 8mm diameter are considered for simulation. Sub-channel analysis clarifies that coolant flow distribution becomes uniform when the gap width is set to 1.0 mm. was less than 620°C. Effects of various parameters, such as boundary conditions and pitch-to-diameter ratios, on the mixing phenomenon in sub-channels and heat transfer are investigated. The effect of pitch-to-diameter ratio (P/D) on the distributions of surface temperature and heat transfer coefficient (HTC) in a sub-channel, it was found that HTC increases with P/D 1.1 first and then decreases with increasing P/D ratio. Apart from the basic geometry sub-channel, a square sub-channel with a wire-wrapped rod inside has been chosen to investigate the “wire effect”.

Prediction of the Nuclear Fuel Rod Abrasion in the Probability Sense

2016 ◽  
Vol 821 ◽  
pp. 317-324
Author(s):  
Vladimír Zeman ◽  
Zdeněk Hlaváč
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Fretting Wear ◽  
Primary Circuit ◽  
Grid Cells ◽  
Operational Parameters ◽  
Spacer Grid ◽  
Mean Values ◽  
Fuel Rod ◽  
Lower Limits

The paper deals with the upper and lower limits estimation of the friction work and fretting wear in the contact of nuclear fuel rods with fuel assembly (FA) spacer grid cells. The friction work is deciding factor for the prediction of the fuel rod cladding abrasion caused by FA vibration. Design and operational parameters of the FA components are understood as random variables defined by mean values and standard deviations. The gradient and three sigma criterion approach is applied to the calculation of the upper and lower limits of the friction work and fretting wear in particular contact surfaces between the fuel rod cladding and some of spacer grid cells. The fuel assembly vibration is excited by pressure pulsations of the cooling liquid generated by main circulation pumps in the coolant loops of the NPP primary circuit. The method is applied for hexagonal type nuclear fuel assembly in the VVER type reactors.

Investigations on the Movement of Fuel Elements in the Core of a Pebble Bed Reactor

1969 ◽  
Vol 91 (2) ◽  
pp. 390-394
Author(s):  
D. Bedenig ◽  
C. B. v. d. Decken ◽  
W. Rausch
Keyword(s):  
Theoretical Model ◽  
Flow Behavior ◽  
Experimental Studies ◽  
Experimental Equipment ◽  
Pebble Bed ◽  
Pebble Bed Reactor ◽  
The Core ◽  
Method Of Measurement ◽  
Fuel Elements ◽  
Type Fuel

For several years gas-cooled high temperature reactors have been developed in Germany, the main feature of which are their pebble-type fuel elements. The pebble bed is in the state of a continuous circulation process which is the reason for a series of nuclear and technical advantages. To make use of these advantages, comprehensive experimental studies on the flow behavior of a pebble bed were carried out. First, experimental equipment and the most successful method of measurement are described. Then typical results of parameter studies are reported as well as a theoretical model to calculate the pebble bed flow behavior. At last typical functions describing the flow behavior in the core of the THTR 300 MWe Prototype Reactor are reported.

CFD Analysis of Boron Distribution at Core Entrance in Case of Reverse Flow Induced From PRISE

2014 ◽  
Author(s):  
Svetlin Philipov
Keyword(s):  
Boric Acid ◽  
Input Data ◽  
Current Issue ◽  
Reverse Flow ◽  
Primary Circuit ◽  
Cfd Analysis ◽  
Reversed Flow ◽  
The Core ◽  
Loss Of Coolant ◽  
Boron Distribution

Initiating events such as primary to secondary loss of coolant (PRISE) can lead to conditions forming reversed flow from the second to the primary circuit. Current issue shows the results of a CFD analysis of the distribution of boric acid on the entrance of the core in case of such reversed flow of coolant as a result of PRISE initiation event. Analyzed accident is included in the list of design basis accidents and requires precise approach in analyzing the phenomena associated with the possibility of injection of coolant with low concentration of boric acid in the primary side. The paper emphasizes on the application of CFD to solve the problem. Analyzing the accident is done in advance with the help of system code RELAP. The input data as flow rate, concentration and temperature at the inlet of the reactor is submitted as boundary conditions in FLUENT and boric acid mixing is analyzed to the core inlet.

Numerical Simulation of the Behavior of Fission Products in the Primary Circuit of the VVER During the LOCA Severe Accident

2009 ◽  
Author(s):  
S. V. Tsaun ◽  
V. V. Bezlepkin ◽  
A. E. Kiselev ◽  
I. A. Potapov ◽  
V. F. Strizhov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Fission Products ◽  
Severe Accident ◽  
Primary Circuit ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Loss Of Coolant ◽  
Mass Activity

The methods and models for the analysis of the radiological consequences of the design basis and severe accidents in a Nuclear Power Plant (NPP) are presented in this paper when using the system code SOCRAT. The system code SOCRAT/V3 was elaborated for a realistic analysis of radiological consequences of severe accidents in a NPP. The following models of the fission products (FP) behavior are included into the code SOCRAT/V3: (i) the condensation and the evaporation of the FP in the gaseous phase and (ii) the sedimentation, the evaporation, and the coagulation of the aerosol-shape FP. The latter processes are governed by gravity, Brownian and turbulent diffusion, thermophoresis, turbophoresis and so forth. The behavior of the FP during the loss-of-coolant accidents (LOCA) is presented to demonstrate the possibilities of the code SOCRAT/V3. The main stages of the accident (the core dryout, the core reflooding, the core degradation, the hydrogen generation, the FP release, etc.) are described. Corresponding estimations of the mass, activity, and decay heat of the suspended, settled and released into containment the FP (Xe, Te, Cs, CsI, Cs2MoO4, and so forth) are represent as well.

Development of Real-Time Thermal-Hydraulic Analysis Code for Plate Type Fuel Reactor

2010 ◽  
Author(s):  
Lei Li ◽  
Zhijian Zhang
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Real Time ◽  
Transient Flow ◽  
Rectangular Channel ◽  
Hydraulic Analysis ◽  
Fuel Reactor ◽  
Thermal Hydraulic Analysis ◽  
Type Fuel ◽  
Plate Type

A multi-channel model thermal-hydraulic analysis code in real-time for plate type fuel reactor is developed in this paper. In this code, every fuel assembly in reactor is divided into a subchannel. A series of reasonable mathematical and physical model are set up based on the structure and operational characteristics of plate type fuel core. As for the choice of flow friction and heat transfer models, all possible flow regimes which include the laminar flow, transient flow and turbulent flow, and heat transfer regimes which include single liquid phase heat transfer, sub-cooled boiling, saturation boiling, film boiling and single vapor phase heat transfer, are considered. The correlations and constitutive equations used in the code are fit for the rectangular channel. Look-up table method is used to calculate the properties of water and steam. The code has been loaded on the real-time simulation supporting system SimExec. The reactivity insertion accident and loss of flow accident, which has been defined in the IAEA 10MW MTR benchmark program, were calculated by the code in this paper for validation. Furthermore, the steady state of CARR (China Advanced Research Reactor) is analyzed by this code. The detailed flow distribution in each fuel assembly is obtained. The temperature of coolant, quality, void fraction, DNBR in each subchannel is calculated. The results show that the recently developed code can be used for real time thermal hydraulic analysis of plate type fuel reactor.

Numerical Simulation of the Transient Phase Change Heat Transfer in a Pressurized Water Reactor Core During a Loss-of-Coolant Accident

2011 ◽  
Author(s):  
Soo W. Jo ◽  
Yong K. Lee ◽  
Jong C. Jo
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Transient Heat Transfer ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

Temperature of pressurized water reactor (PWR) core is a key parameter used widely for judging the initiation of emergency operating procedures and severe accident management. Since direct measurement of the fuel cladding surface temperature using thermocouples is not practicable currently, the coolant temperature at the core exit locations is monitored instead. Several experimental researches showed that the CET rise during a loss of coolant accident (LOCA) and its magnitudes were always lower than the actual fuel rod cladding temperature at the same time. In this regard, a theoretical analysis of the transient heat transfer of coolant flow in a PWR core is needed to confirm the findings from the previous experimental works. This paper addresses numerical simulation of the transient boiling-induced multiphase flow through a simplified PWR core model during a LOCA by a commercial computational fluid dynamics (CFD) code. The calculated results are discussed to understand the transient heat transfer mechanism in the core and to provide useful technical information for reactor design and operation.

APROS Analysis of Loviisa Sump Strainer Study

2010 ◽  
Author(s):  
Heikki Kantee ◽  
Harri Kontio
Keyword(s):  
Safety Issue ◽  
Steel Wire ◽  
Life Time ◽  
Simulation Software ◽  
Test Facility ◽  
Primary Circuit ◽  
Spacer Grids ◽  
The Core ◽  
Loss Of Coolant ◽  
Core Cooling

The two Loviisa VVER-440 type reactors were commissioned in 1977 and 1980. The original designed life time of the reactors was 30 years. In 2003 Fortum, the owner and the operator of the Loviisa plant, launched an extensive safety study to prove the authorities that there was not any major safety issue why operating license could not be extended for another 20 years. In 2007 the Ministry of Employment and the Economy of Finland granted 20 and 23 years extension to the operating license for units 1 and 2, respectively. One issue, which needed further investigation, was the core cooling capability during sump circulation; i.e. were the present sump strainers good enough to prevent insulation fiber from not clogging the core coolant flow? Back in the 1990’s the original steel wire type sump strainers were replaced with stronger steel pipe type strainers. Some time later experiments were carried out to find out if insulation fiber could penetrate through the strainer holes and reduce the coolant mass flow rate through the core. The experiments indicated that the insulation fiber mixed with coolant partly penetrates through the strainer and gathers to the fuel assembly spacer grids increasing pressure loss across the core. The experiments were carried out in a rather simple test facility and also under forced single phase circulation. In those loss-of-coolant accidents (LOCA) where sump circulation takes place, circumstances are completely different. Therefore, it was decided that the APROS (Advanced PROcess Simulation) simulation software would be used to study the insulation fiber effect on core coolability during the accident. A large LOCA was chosen for the case to be analyzed. The reason for this was that during a large LOCA sump circulation begins in the early phase of the accident and a lot of emergency core cooling (ECC) water is injected into the primary circuit during sump circulation. The paper will first shortly discuss APROS simulation software. Then the test facility and the experimental results will be presented. The main issue is the analyses results. Several analyses were carried out to be able to determine the maximum amount fiber gathered in the spacer grids which the core can tolerate without overheating.

scholarly journals Numerical Investigation of Conjugated Heat Transfer of the Plate-Type Fuel Assembly in the Research Reactor

2021 ◽  
Vol 9 ◽  
Author(s):  
Quan Li ◽  
Qiang Ma ◽  
Yuanming Li ◽  
Ping Chen ◽  
Chao Ma ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Power Density ◽  
Research Reactor ◽  
Flow Boiling ◽  
Fluid Velocity ◽  
Rensselaer Polytechnic Institute ◽  
Conjugated Heat Transfer ◽  
Type Fuel ◽  
Plate Type

In nuclear reactors, the research of conjugated heat transfer between the fuel and coolant in the fuel assembly is fundamental for improving the safety, reliability and economy. The numerical approach based on Computational Fluid Dynamics (CFD) can be used to realize the rapid analysis of the conjugated heat transfer. Besides, the numerical simulation can provide detailed physical fields that are useful for the designing and optimizing of the fuel assembly. The plate-type fuels are generally used to enhance heat transfer in research reactors with high power density. In this study, a standard plate-type fuel assembly in the research reactor was taken into consideration. The solid-fluid conjugated heat transfer of the fuel assembly and coolant was numerically investigated. In the fluid region, the subcooled flow boiling simulation model was established by implementing the Rensselaer Polytechnic Institute model into the Euler multi-phase flow method. The results show that the conjugated heat transfer of the fuel assembly and coolant can be simulated using the model established in this work. The influence of fluid velocity, power density and the width of the flow channel on the temperature distribution and the conjugated heat transfer was investigated and discussed.

scholarly journals Detection of the Departure from Nucleate Boiling in Nuclear Fuel Rod Simulators

2013 ◽  
Vol 2013 ◽  
pp. 1-7 ◽  
Author(s):  
Amir Zacarias Mesquita ◽  
Rogério Rivail Rodrigues
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Electric Current ◽  
Nuclear Fuel ◽  
Critical Heat Flux ◽  
Cooling Water ◽  
Nucleate Boiling ◽  
Inlet Temperature ◽  
Fuel Rod ◽  
The Core

In the thermal hydraulic experiments to determin parameters of heat transfer where fuel rod simulators are heated by electric current, the preservation of the simulators is essential when the heat flux goes to the critical point. One of the most important limits in the design of cooling water reactors is the condition in which the heat transfer coefficient by boiling in the core deteriorates itself. The heat flux just before deterioration is denominated critical heat flux (CHF). At this time, the small increase in heat flux or in the refrigerant inlet temperature at the core, or the small decrease in the inlet flux of cooling, results in changes in the heat transfer mechanism. This causes increases in the surface temperature of the fuel elements causing failures at the fuel (burnout). This paper describes the experiments conducted to detect critical heat flux in nuclear fuel element simulators carried out in the thermal-hydraulic laboratory of Nuclear Technology Development Centre (CDTN). It is concluded that the use of displacement transducer is the most efficient technique for detecting critical heat flux in nuclear simulators heated by electric current in open pool.

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