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.

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 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.

CFD Analysis of a Unidirectional Axial Turbine for Twin Turbine Topology in OWC Plants

2013 ◽  
Author(s):  
Bruno Pereiras ◽  
Pablo Valdez ◽  
Francisco Castro ◽  
Julio C. Garrido
Keyword(s):  
Numerical Model ◽  
Flow Field ◽  
Reverse Flow ◽  
Cfd Analysis ◽  
Ocean Energy ◽  
Loss Distribution ◽  
Axial Turbine ◽  
Guide Vanes ◽  
The Core ◽  
Reverse Mode

OWC devices are widely known among researchers in ocean energy. It is well-known that the efficiency of the device is closely related to the efficiency of the Power-Take-Off (PTO) which is usually a turbine. Traditionally, self-rectifying turbines are the most widely considered for working in an OWC because unidirectional turbines require a system of valves to rectify the flow. However, another option recently proposed is the use of the “twin turbine” configuration. This paper focuses on the performance of the turbines used in this configuration. A numerical model has been developed and validated with data from the bibliography. This model has been used to analyze the flow field of the turbine when working in both performance modes: direct and reverse. Flow angles and loss distribution have been analyzed and interesting conclusions can be extracted. Once the flow field has been analyzed, changes in the turbine geometry are proposed in order to improve the efficiency of the whole system by increasing the blockage made by the turbine in reverse mode. These changes, focused on the solidity of the rotor and guide vanes, were implemented and new simulations were carried out. The results obtained are the core of this work.

Numerical Simulation of the Insulation Material Transport to a PWR Core Under Loss of Coolant Accident Conditions

2010 ◽  
Author(s):  
Thomas Ho¨hne ◽  
Alexander Grahn ◽  
So¨ren Kliem ◽  
Ulrich Rohde ◽  
Frank-Peter Weiss
Keyword(s):  
Liquid Velocity ◽  
Reverse Flow ◽  
Mineral Wool ◽  
Insulation Material ◽  
Material Transport ◽  
Spacer Grids ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

In 1992, strainers on the suction side of the ECCS pumps in Barseba¨ck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally-insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modelled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the insulation material during reverse flow. This will certainly further improve the coolability of the core. The spacer grids were modelled as a strainer, which completely retains all the insulation material reaching the uppermost spacer level. There, the accumulation of the insulation material gives rise to the formation of a compressible fibrous cake, the permeability of which to the coolant flow is calculated in terms of the local amount of deposited material and the local value of the superficial liquid velocity. Before the switch over of the ECC injection from the flooding mode to the sump mode, the coolant circulates in an inner convection loop in the core extending from the lower plenum to the upper plenum. The CFD simulations have shown that after starting the sump mode, the ECC water injected through the hot legs flows down into the core at so-called “breakthrough channels” located at the outer core region where the downward leg of the convection roll had established. The hotter, lighter coolant rises in the centre of the core. As a consequence, the insulation material is preferably deposited at the uppermost spacer grids positioned in the breakthrough zones. This means that the fibers are not uniformly deposited over the core cross section. When the inner recirculation stops later in the transient, insulation material can also be collected in other regions of the core. Nevertheless, with a total of 2.7 kg fiber material deposited at the uppermost spacer level, the pressure drop over the fiber cake is not higher than 8 kPa and all the ECC water could still enter the core.

A window detection technique with adjustable threshold for transmitted reference receivers

2013 ◽  
Vol 3 (1) ◽  
Author(s):  
Apratim Roy ◽  
A. Rashid
Keyword(s):  
Input Data ◽  
Cmos Technology ◽  
Ultra Wideband ◽  
Power Requirement ◽  
The Core ◽  
Transmitted Reference ◽  
Detection Circuit ◽  
Detection Window ◽  
Improved Performance ◽  
Decision Circuit

AbstractThis paper presents a threshold decision circuit with an adjustable detection window designed in a 90-nm IBM CMOS technology. Together with an RF mixer, the decision Section realizes the circuit implementation of the back-end of a transmitted reference ultra wideband receiver, which is yet to be reported in literature. The proposed circuit is built on a differential amplifier core and avoids the use of integrator and sampling blocks, which reduces the device burden necessary for the architecture. Moreover, the detection window threshold of the design can be regulated by three independent factors defined by the circuit elements. The circuit is tested at an input data rate of 0.1∼2.0 Gbps and the core decision section consumes 9.14 mW from a 1.2-V bias supply (with a maximum capacity/Pdc ratio of 218.8 GHz/W). When compared against other reported decision blocks, the proposed detection circuit shows improved performance in terms of capacity and power requirement.

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

Examination of Dust in AVR Pipe Components

2008 ◽  
Author(s):  
Johannes Fachinger ◽  
Heiko Barnert ◽  
Alexander P. Kummer ◽  
Guido Caspary ◽  
Manuel Seubert ◽  
...  
Keyword(s):  
Particle Size ◽  
Optical Microscopy ◽  
Phase 1 ◽  
Median Number ◽  
Primary Circuit ◽  
Pipe Surface ◽  
Dose Rates ◽  
The Core ◽  
Chemical Decontamination ◽  
Dust Surface

Pebble Bed HTGR’s like the AVR in Ju¨lich have the advantage of continuous fuelling. However the multiple passes of the fuel pebbles through the core have the disadvantage that the pebble’s movement through the fuelling system and the core produces graphite dust. This dust is transported from the core to other parts of the primary circuit and deposits on components. Although previous experiments performed during AVR operation have given some insight into the dust particle size and activity, there is little information on the behaviour of the dust that was deposited in the system. The decommissioning of the AVR has provided the opportunity to sample and characterise such dust from a number of components and gauge the adhesion strength. From the side of PBMR Pty Ltd this opportunity is considered important to enhance the knowledge about dust characteristics before the PBMR Demonstration Power Plant (DPP) is operational and able to produce specific plant information through sampling and analysis. AVR GmbH has provided a number of pipes and joints for investigation of loose and bound dust. Phase 1 of the analysis was used to determine the best techniques to be used on larger items. No measurable loose dust could be collected. Thereupon rings were cut from a T-section and subdivided into eight segments. The surface of the untreated segments were photographed and documented by optical microscopy, the dose rates were measured and gamma-spectrometry performed. Following this a mechanical or chemical decontamination was carried out to remove and isolate the bound dust. The average isolated dust amount was about 2 mg/cm2. Both decontamination processes indicates a strong bonding of the dust surface layer. In the case of mechanical decontamination about 60% and by chemical decontamination about 95% of the radionuclide inventory could be removed. The contribution of removed metal needs to be investigated in more detail. The median number related particle size measured by optical microscopy was found to be in the range of 0.2 to 0.7 μm whereas the median weight related size is in the range of 0.8 to 1.5 μm. The initial results indicate that this dust sticks very strongly to the pipe surface. Phase 2 will concentrate on longer pieces of piping where hopefully more loose dust can be obtained and analysed. If the same strong bonding is observed the reason for this phenomenon needs to be explained and perhaps tested with non-active dust.

scholarly journals Analysis of mass transfer processes in a reactor during a loss-of-coolant accident

2018 ◽  
Vol 4 (2) ◽  
pp. 149-154
Author(s):  
Aleksey Kulikov ◽  
Andrey Lepyokhin ◽  
Vitaly Polunichev
Keyword(s):  
Time Interval ◽  
Initial Water ◽  
Pressurized Water ◽  
The Core ◽  
Maximum Time ◽  
Loss Of Coolant Accident ◽  
Safe State ◽  
Loss Of Coolant ◽  
Water Gas ◽  
Gas Ratio

The purpose of the work was to optimize the parameters of the spillage system equipped with a gas pressure hydroaccumulator for a ship pressurized water reactor in a loss-of-coolant accident. The water-gas ratio in the hydroaccumulator and the hydraulic resistance of the path between the hydroaccumulator and the reactor were optimized at the designed hydroaccumulator geometric volume. The main dynamic processes were described using a mathematical model and a computational analysis. A series of numerical calculations were realized to simulate the behavior dynamics of the coolant level in the reactor during the accident – by varying the optimized parameters. Estimates of the minimum and maximum values of the coolant level were obtained: depending on the initial water-gas ratio in the hydroaccumulator at different diameters of the flow restrictor on the path between the hydroaccumulator and the reactor. These results were obtained subject to the restrictive conditions that, during spillage, the coolant level should remain above the core and below the blowdown nozzle. The first condition implies that the core is in safe state, the second excludes the coolant water blowdown. The optimization goal was to achieve the maximum time interval in which these conditions would be satisfied simultaneously. The authors propose methods for selecting the optimal spillage system parameters; these methods provide the maximum time for the core to be in a safe state during a loss-of-coolant accident at the designed hydroaccumulator volume. Using these methods, it is also possible to make assessments from the early stages of designing reactor plants.

scholarly journals Thermal Hydraulic Analysis of a Nuclear Reactor due to Loss of Coolant Accident with and without Emergency Core Cooling System

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Author(s):  
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari
Keyword(s):  
Nuclear Reactor ◽  
Cooling System ◽  
Computational Study ◽  
Reactor Core ◽  
Primary Circuit ◽  
Feed Water ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling

This paper evaluates the thermal hydraulic behavior of a pressurized water reactor (PWR) when subjected to the event of Loss of Coolant Accident (LOCA) in any channel surrounding the core. The accidental break in a nuclear reactor may occur to circulation pipe in the main coolant system in a form of small fracture or equivalent double-ended rupture of largest pipe connected to primary circuit line resulting potential threat to other systems, causing pressure difference between internal parts, unwanted core shut down, explosion and radioactivity release into environment. In this computational study, LOCA for generation III+ VVER-1200 reactor has been carried out for arbitrary break at cold leg section with and without Emergency Core Cooling System (ECCS). PCTRAN, a thermal hydraulic model-based software developed using real data and computational approach incorporating reactor physics and control system was employed in this study. The software enables to test the consequences related to reactor core operations by monitoring different operating variables in the system control bar. Two types of analysis were performed -500% area break at cold leg pipe due to small break LOCA caused by malfunction of the system with and without availability of ECCS. Thermal hydraulic parameters like, coolant dynamics, heat transfer, reactor pressure, critical heat flux, temperature distribution in different sections of reactor core have also been investigated in the simulation. The flow in the reactor cooling system, steam generators steam with feed-water flow, coolant steam flow through leak level of water in different section, power distribution in core and turbine were plotted to analyze their behavior during the operations. The simulation showed that, LOCA with unavailability of Emergency Core Cooling System (ECCS) resulted in core meltdown and release of radioactivity after a specific time.

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