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.

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.  

2018 ◽  
pp. 3-10
Yu. Kovbasenko ◽  
Yevgen Bilodid

The article investigates the possibility of a self-sustaining chain nuclear fission reaction during the development of a severe accident in the core at nuclear power plants with reactors WWER-1000 of Ukraine. Some models for calculating a criticality at different stages of the severe accident in the reactor VVER-1000 vessel were developed and calculations of multiplication properties of fuel containing masses were performed. The severe accident in the VVER-1000 core approximately divided into seven major stages: the intact reactor core, beginning of cladding damage (swelling), cladding melting and flowing down to the support grid, melting of constructional materials, homogenization of the materials at the bottom of the reactor vessel, stratification of corium at the bottom of the reactor vessel, the exit of the corium from the reactor shaft. It was shown that at the beginning of an accident, if fuel rods geometry is maintained, criticality might appear even if the emergency protection rods is triggered. With further development of the accident, the melt of fuel and structural materials will be deeply subcritical if water cannot penetrate into the pores or voids of the melt. In the case of the formation of pores or voids in the melt and the ingress of water into them, a recriticality may arise. A compensating measure is the addition of a boric acid solution to a cooling water with a certain concentration. According to the results of the computation analysis, a reactor core loaded with TVSA fuel (Russian production) requires a higher concentration of boric acid in water to compensate the multiplication properties of the fuel system in emergency situations compared to the core loaded with TVS-WR fuel (manufactured by Westinghouse), i.e. TVS-WR fuel is safer from the criticality point of view.

2016 ◽  
Vol 301 ◽  
pp. 59-73 ◽  
A. Rajamani ◽  
T. Sundararajan ◽  
B.V.S.S.S. Prasad ◽  
U. Parthasarathy ◽  
K. Velusamy

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

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.

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).

Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

2021 ◽  
Vol 23 (2) ◽  
pp. 63
Muhammad Budi Setiawan ◽  
Pande Made Udiyani

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach and a graded approach. The normal source term is calculated assuming the presence of impurities on the fuel plate, due to fabrication limitations. Meanwhile, the abnormal source term is postulated in the LOCA event. The core reactor inventory and source term is divided into 8 radionuclide groups which are Noble gasses group (Xe, Kr); Halogen (I); Akali Metal (Cs, Rb); Tellurium Group (Te, Sb, Sc); Barium-Strontium Group (Ba, Sr); Noble Metals (Ru, Rh, Pd, Mo, Tc, Co); Lanthanides group (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) and Cerium Group (Ce, Pu , Np).

2018 ◽  
Vol 2018 ◽  
pp. 1-11
Jiarun Mao ◽  
Lei Song ◽  
Yuhao Liu ◽  
Jiming Lin ◽  
Shanfang Huang ◽  

This paper presents capacity of the passive decay heat removal system (DHRS) operated under the natural circulation conditions to remove decay heat inside the main vessel of the Lead-bismuth eutectic cooled Fast Reactor (LFR). The motivation of this research is to improve the inherent safety of the LFR based on the China Accelerator Driven System (ADS) engineering project. Usually the plant is damaged due to the failure of the main pumps and the main heat exchangers under the Station Blackout (SBO). To prevent this accident, we proposed the DHRS based on the diathermic oil cooling for the LFR. The behavior of the DHRS and the plant was simulated using the CFD code STAR CCM+ using LFR with DHRS. The purpose of this analysis is to evaluate the heat exchange capacity of the DHRS and is to provide the reference for structural improvement and experimental design. The results show that the stable natural circulations are established in both the main vessel and the DHRS. During the decay process, the heat exchange power is above the core decay heat power. In addition, in-core decay heat and heat storage inside the main vessel are efficiently removed. All the thermal-hydraulics parameters are within a safe range. Moreover, the highest temperature occurs at the upper surface of the core. A swirl occurs at the corner of the lateral core surface and some improvements should be considered. And the natural circulation driving force can be further increased by reducing the loop resistance or increasing the natural circulation height based on the present design scenario to enhance the heat exchange effect.

2017 ◽  
Vol 743 ◽  
pp. 389-393 ◽  
Andrei V. Morozov ◽  
Anna V. Pityk ◽  
Sergei V. Ragulin ◽  
Aleksandra S. Soshkina

The results of hand calculation of boric acid accumulation in the core in a new generation WWER-TOI reactor in case of LOCA are presented. Variants of reducing the H3BO3 concentration in the HA-3 system down to 1, 2, 4 and 8 g/kg are considered. The mass of boric acid deposits on the core internals depending on the value of boric acid concentration in the HA-3 system is determined. The obtained results allow concluding that the accumulation and crystallization of boric acid in the core is possible.

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