scholarly journals Effect of Boric Acid Solubility in Steam on the Process of Mass Transfer during Emergency Cooling of VVER-1200 Nuclear Reactor

Md Rezouanul Kabir ◽  
Morozov A.V. ◽  
Md Saif Kabir

The mechanisms of boric acid mass transfer in a VVER-1200 reactor core are studied in this work in the event of a major circulatory pipeline rupture and loss of all AC power. The VVER-1200's passive core cooling technology is made up of two levels of hydro accumulators. They use boric acid solution with a concentration of 16 g H3BO3/kg H2O to control the reactivity. Because of the long duration of the accident process, the coolant with high boron content starts boiling and steam with low concentration of boric acid departs the core. So, conditions could arise in the reactor for possible accumulation and subsequent crystallization of boric acid, causing the core heat removal process to deteriorate. Calculations were carried out to estimate the likelihood of H3BO3 build-up and subsequent crystallization in the core of the VVER reactor. According to the calculations, during emergency the boric acid concentration in the reactor core is 0.153 kg/ kg and 0.158 kg/kg in both the events of solubility of steam and without solubility of steam respectively and it does not exceed the solubility limit which is about 0.415 kg/kg at water saturation temperature. No precipitation of boric acid occurs within this time during the whole emergency process. Therefore, findings of the study can be used to verify whether the process of decay heat removal is affected or not.

2014 ◽  
Vol 986-987 ◽  
pp. 231-234
Jun Teng Liu ◽  
Qi Cai ◽  
Xia Xin Cao

This paper regarded CNP1000 power plant system as the research object, which is the second-generation half Nuclear Reactor System in our country, and tried to set Westinghouse AP1000 passive residual heat removal system to the primary circuit of CNP1000. Then set up a simulation model based on RELAP5/MOD3.2 program to calculate and analyze the response and operating characteristic of passive residual heat removal system on assumption that Station Blackout occurs. The calculation has the following conclusions: natural circulation was quickly established after accident, which removes core residual heat effectively and keep the core safe. The residual heat can be quickly removed, and during this process the actual temperature was lower than saturation temperature in reactor core.

2017 ◽  
Vol 39 (4) ◽  
pp. 55-60
A. A. Avramenko ◽  
N. P. Dmitrenko ◽  
М. M. Kovetskaya ◽  
Yu. Yu. Kovetskaya

Heat and mass transfer in a model of the core of a nuclear reactor with spherical fuel elements and a helium coolant was studied. The effect of permeability of the pebble bed zone and geometric parameters on the temperature distribution of the coolant in the reactor core is analyzed.  

2021 ◽  
Vol 7 (4) ◽  
pp. 291-295
Denis A. Pakholik ◽  
Oleg Yu. Kochnov ◽  
Valery V. Kolesov ◽  
Vladimir V. Fomichev

There are various ways to obtain Mo-99. Some of them are widely used in industrial production, others are in the research stage with the aim of increasing the product yield. The main industrial method for obtaining Mo-99 using a nuclear reactor is the fragmentation method. This method provides for the presence of a uranium target and a nuclear reactor. The target is placed in the channel of the reactor core and irradiated with neutrons for the required time. After that, the target is removed from the channel to the “hot” chamber for the chemical separation of Mo-99. This is how Mo-99 is obtained practically all over the world. The paper considers the fragmentation method for producing Mo-99, which is implemented on the basis of the engineering and technological complex of the VVR-c research nuclear reactor. In order to increase the yield of Mo-99, a modernized model of the “tube-in-tube” target is proposed. The assessment of the production of Mo-99 and the cooling efficiency of the modernized target was carried out. The calculations were performed using the VisualBurnOut and Ansys CFX software packages. Computational studies have shown an increase in the energy release and the amount of the produced Mo-99 isotope in the target of the modernized design. In the most stressed zones, the target wall temperature exceeds the water saturation temperature. Surface boiling occurs in these zones. As a result, turbulization and mixing of the near-wall boundary water layer increases. This improves heat dissipation.

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev

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.

Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

2019 ◽  
Vol 5 (1) ◽  
pp. 75-80
Vyacheslav S. Kuzevanov ◽  
Sergey K. Podgorny

The need to shape reactor cores in terms of coolant flow distributions arises due to the requirements for temperature fields in the core elements (Safety guide No. NS-G-1.12. 2005, IAEA nuclear energy series No. NP-T-2.9. 2014, Specific safety requirements No. SSR-2/1 (Rev.1) 2014). However, any reactor core shaping inevitably leads to an increase in the core pressure drop and power consumption to ensure the primary coolant circulation. This naturally makes it necessary to select a shaping principle (condition) and install heat exchange intensifiers to meet the safety requirements at the lowest power consumption for the coolant pumping. The result of shaping a nuclear reactor core with identical cooling channels can be predicted at a quality level without detailed calculations. Therefore, it is not normally difficult to select a shaping principle in this case, and detailed calculations are required only where local heat exchange intensifiers are installed. The situation is different if a core has cooling channels of different geometries. In this case, it will be unavoidable to make a detailed calculation of the effects of shaping and heat transfer intensifiers on changes in temperature fields. The aim of this paper is to determine changes in the maximum wall temperatures in cooling channels of high-temperature gas-cooled reactors using the combined effects of shaped coolant mass flows and heat exchange intensifiers installed into the channels. Various shaping conditions have been considered. The authors present the calculated dependences and the procedure for determining the thermal coolant parameters and maximum temperatures of heat exchange surface walls in a system of parallel cooling channels. Variant calculations of the GT-MHR core (NRC project No. 716 2002, Vasyaev et al. 2001, Neylan et al. 1994) with cooling channels of different diameters were carried out. Distributions of coolant flows and temperatures in cooling channels under various shaping conditions were determined using local resistances and heat exchange intensifiers. Preferred options were identified that provide the lowest maximum wall temperature of the most heat-stressed channel at the lowest core pressure drop. The calculation procedure was verified by direct comparison of the results calculated by the proposed algorithm with the CFD simulation results (ANSYS Fluent User’s Guide 2016, ANSYS Fluent. Customization Manual 2016, ANSYS Fluent. Theory Guide 2016, Shaw1992, Anderson et al. 2009, Petrila and Trif 2005, Mohammadi and Pironneau 1994).

2020 ◽  
Vol 01 (02) ◽  
pp. 53-60
Pronob Deb Nath ◽  
Kazi Mostafijur Rahman ◽  
Md. Abdullah Al Bari

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.

2013 ◽  
Vol 15 (4) ◽  
pp. 33
Anwar Ilmar Ramadhan ◽  
Indra Setiawan ◽  
M. Ivan Satryo

Safety is an issue that is of considerable concern in the design, operation and development of a nuclear reactor. Therefore, the method of analysis used in all these activities should be thorough and reliable so as to predict a wide range of operating conditions of the reactor, both under normal operating conditions and in the event of an accident. Performance of heat transfer to the cooling of nuclear fuel, reactor safety is key. Poor heat removal performance would threaten the integrity of the fuel cladding which could further impact on the release of radioactive substances into the environment in an uncontrolled manner to endanger the safety of the reactor workers, the general public, and the environment. This study has the objective is to know is profile contour of fluid flow and the temperature distribution pattern of the cooling fluid is water (H2O) in convection in to SMR reactor with fuel sub reed arrangement of hexagonal in forced convection. In this study will be conducted simulations on the SMR reactor core used sub channel hexagonal using CFD (Computational Fluid Dynamics) code. And the results of this simulation look more upward (vector of fluid flow) fluid temperature will be warm because the heat moves from the wall to the fluid heater. Axial direction and also looks more fluid away from the heating element temperature will be lower.

2013 ◽  
Vol 448-453 ◽  
pp. 1907-1911
Wei Zhi Jia ◽  
Rui Wang ◽  
Yun Zhou

As the core monitoring system of AP1000, BEACON always uses a full-core nodal model for core monitoring based on the ANC-NEM nodal model. The theory behind the nodal expansion method is discussed, and the application of the method in BEACON is described. Finally, an ANC-NEM calculation simulation is proposed.

1970 ◽  
Vol 10 (04) ◽  
pp. 337-348 ◽  
F.I. Stalkup

Abstract Displacements of laboratory oils by propane in long, consolidated sandstone cores in the presence of high water saturations have shown that oil recoveries approaching 100 percent may be realized by continuous water-propane injection, even for oil saturations close to residual oil. However, it was often necessary to inject many pore volumes of solvent to attain this high a recovery. Initial oil saturations were established by injecting water and oil at a constant ratio into the porous medium containing residual oil to a waterflood until a steady state was obtained. Propane and water were then injected in the same fixed ratio to displace the oil. These and other experiments indicate that in the presence of a high water saturation only part of the presence of a high water saturation only part of the oil is flowable. Part resides in locations that are blocked by water, and the oil in these stagnant locations is not flowable. This nonflowable oil, it is believed, can be recovered by molecular diffusion into the flowing propane of a water-propane displacement. Values for the saturation of hydrocarbon that is contained in the stagnant locations and values for the ratio of the longitudinal hydrodynamic-dispersion coefficient to displacement velocity were determined at various water saturations in the test sandstones. The data suggest that rock wettability may influence the stagnant saturation and that stagnant oil saturations may not be as large in reservoir rocks as they are observed to be in laboratory sandstones. Mass transfer between the flowing solvent and hydrocarbon components in the stagnant saturation was expressed by a first-order rate expression. Rough values for the mass transfer coefficients for the propane-trimethylhexane hydrocarbon pair were estimated from experiments. Computations using these values for mass transfer coefficients indicate that experiments in laboratory-size cores may show much poorer displacement efficiency than that which might actually occur in the field. Introduction Injection of water with light hydrocarbon solvents is a technique that may be used to partially control solvent mobility. The higher water saturation forced by water injection reduces the permeability to solvent flow, and the mobility of the solvent region is reduced relative to that of the oil-bank region. However, it also might be expected that this higher water saturation influences the microscopic unit displacement of oil by solvent to some degree. For example, as discussed by Thomas et al., two possible effects of high water saturation on the displacement mechanism come to mind. First, a miscible displacement in the presence of water is operating on a different pore-size distribution than if no water were present. Pore-size distribution and the dp term (product of the microscopic inhomogeneity factor and "effective" particle diameter) may considerably influence the magnitudes of transverse and longitudinal dispersion coefficients. Secondly, in a multiphase system the wetting phase may trap single pores or even isolate large fingers or dendrites of the nonwetting phase. The nonwetting phase in these dead-end pores or dendrites would be phase in these dead-end pores or dendrites would be nonflowing and might either be completely isolated by the wetting phase or might communicate with the flowing nonwetting fluid by diffusion through nonwetting fluid-filled pores. Aspects of miscible displacement in the presence of water have been investigated by a number of researchers. Fitzgerald and Nielson observed that the simultaneous injection of gasoline and water into a Berea sandstone core in a 1:2 ratio recovered only 36 percent of the Bradford crude oil left in the core after waterflooding, and that only 55 to 75 percent recoveries were obtained for simultaneous water-solvent injection into the core when it contained crude oil at connate water saturation. Moreover, these authors reported recoveries of only 60 to 80 percent when solvent alone was injected into the core to displace residual oil to a waterflood. Raimondi et al. injected ethyl benzene (oil) and water simultaneously into a Berea sandstone core to establish flowing oil and water saturations and then injected heptane (solvent) and water simultaneously into the core to miscibly displace the ethyl benzene. SPEJ p. 337

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