Model & data based assessment of the impacts of drawdown of the Chornobyl NPP Cooling Pond on the Cs-137 concentrations in water, sediments and biota

Author(s):  
Roman Bezhenar ◽  
Mark Zheleznyak ◽  
Dmitri Gudkov ◽  
Volodymyr Kanivets ◽  
Gennady Laptev ◽  
...  

<p>Cooling Pond (CP) of the Chornobyl Nuclear Power Plant (ChNPP) is one of the most radioactively contaminated large water bodies over the globe. During the active phase of the ChNPP accident, radionuclides got into the CP in result of atmospheric deposition, release of highly contaminated water from system of accidental cooling, and water used to extinguish the fire. In the years after the accident, the contamination was distributed in the CP due to currents. For this period, three types of hydrological conditions dominated in the CP. Initially, the currents were forced by the cooling system of the ChNPP, which caused a circular movement of water. After the decommissioning of the ChNPP, the natural circulation took place in the CP. Starting from the end of 2014, when pumps that continuously fed the CP with water from the Prypiat River were shutdown, a gradual decrease of water level began. Now the water level has dropped by about 6 m leading to the transformation of the whole reservoir into several small lakes and redistribution of radionuclides in them. The objectives of the study were to calibrate models, which were customized for the CP, using data for the whole post-accident period including data collected during the drawdown period by the joint efforts of Ukrainian and Japanese researchers, and then to provide model based predictions of the future radionuclide concentrations in new water bodies.</p><p>During field studies that were carried out in November 2020, the current state of radioactive contamination of the CP was investigated. Samples of water, suspended and bottom sediments and biota were taken in 9 closed or semi-closed water bodies formed after partial drying of the CP. Concentrations of Cs-137 and its distribution in dissolved and particulated forms were measured in the laboratory. For simulations, the modeling system that consists of the 3D model of thermohydrodynamics and radionuclide transfer THREETOX and the box model POSEIDON-R was created. The THREETOX model was used for the obtaining currents in the CP for each type of hydrological conditions. The POSEIDON-R model was applied for the long-term simulations of the changes of activity concentration in the water, bottom sediments and biota starting from the 1986. The system of boxes in the POSEIDON-R model includes shallow and deep-water boxes. It was built in such a way that after the water level in the CP fell, the calculations were performed only in deep-water boxes. Fluxes of water between boxes were calculated based on currents from the THREETOX model. Seasonal changes in distribution coefficient K<sub>d</sub> describing the partition of Cs-137 concentration between water and sediments were also taken into account. Calculated concentrations of Cs-137 in water and bottom sediments agree well with measurements for all boxes and for entire modeling period. It has been shown that POSEIDON-R model is able to reproduce changes in the concentrations of Cs-137 in freshwater fish occupying different levels the food chain. Scenarios for the potential changes of Cs-137 concentrations were considered by the variation of basic parameters.</p>

Author(s):  
Yu Yu ◽  
Shengfei Wang ◽  
Fenglei Niu

In order to improve the safety of new generation nuclear power plant, passive containment cooling system is innovatively used in AP1000 reactor design. However, since the system operation is based on natural circulation, physical process failure — natural circulation cannot establish or be maintained — becomes one of the important failure modes. Uncertainties in the physical parameters such as heat and cold source temperature and in the structure parameters have important effect on the system reliability. In this paper, thermal–hydraulic model is established for passive containment cooling system in AP1000 and the thermal–hydraulic performance is studied, the effect of factors such as air temperature and reactor power on the system reliability are analyzed.


Author(s):  
Yu Yu ◽  
Shengfei Wang ◽  
Fenglei Niu

Passive containment cooling system (PCCS) is an important safety-related system in AP1000 nuclear power plant, by which heat produced in reactor is transferred to the heat sink – atmosphere – based on natural circulation, independent of human response or the operation of outside equipments, so the reactor capacity of resisting external hazards (earthquake, flood, etc.) is improved. However since the system operation based on natural circulation, many uncertainty factors such as temperatures of cold and heat sources will affect the system reliability, and physical process failure becomes one of the important contributors to system failure, which is not considered in the active system reliability analysis. That is, the system will lose its function since the natural circulation cannot be established or kept even when the equipments in the system can work well. The function of PCCS in AP1000 is to transfer the heat produced in the containment to the environment and to keep the pressure in the containment below its threshold. After accidents the steam is injected to the containment and can be cooled and condensed when it arrives at the containment wall, then the heat is transferred to the atmosphere through the steel vessel. So the peak value of the pressure is influenced by the steam situation which is injected into the containment and the heat transfer and condensate processes under the accidents. In this paper the dynamic thermal-hydraulic (T-H) model simulating the fluid performance in the containment is established, based on which the system reliability model is built. Here the total pressure in the containment is used as the success criteria. Apparently the system physical process failure may be related to the system working state, the outside conditions, the system structure parameters and so on, and it’s a heavy work to analyze the influences of all the factors, so only the effects of important ones are included in the model. Monte Carlo (MC) simulation is used to evaluate the system reliability, in which the input parameters such as air temperature are sampled based on their probabilistic density distributions. The pressure curves along with the accident development are gained and the system reliabilities under different accidents are gotten as well as the main contributors. The results illustrate that the system physical process failure probabilities are varied under different climate conditions, which result in the system reliability and the main contributors to system failure changing, so the different methods can be taken to improve the system reliability according to the local condition of the nuclear power plant.


Author(s):  
Chunhui Dai ◽  
Jun Wu ◽  
Sichao Tan ◽  
Zhenxing Zhao ◽  
Qi Xiao ◽  
...  

Ship nuclear power platform is a small and movable power plant on the sea, aiming at generating electric energy and producing fresh water, it provides support for the national energy strategy. Subsequent to a loss of coolant accident (LOCA), steam is vented in the reactor containment following vaporization of liquid and/or steam expansion. The temperature as well as pressure in the condensation rises synchronously. For removing heat and reducing pressure inside containment subsequent to a LOCA, the Passive containment cooling system of Ship nuclear power platform is designed. In order to establish and maintain the passive heat removing channel, steam condenses on the containment condenser tube surface, coupling natural convection of the seawater inside the tubes. The heat transfer mechanism of Passive containment cooling system is very complex. To solve this problem, a three dimensional heat exchanging/one dimensional natural circulation coupling numerical computing method is proposed to obtained the safety performance of the reactor containment. Models of heat exchanging process between steam which contains non-condensable gas inside the reactor containment and sea water outside are firstly established. Then the thermal-hydraulic characteristics of the steam and sea water beside the heat transfer tubes are obtained by a simulation which is carried out in a LOCA.


Author(s):  
Yu Yu ◽  
Shengfei Wang ◽  
Xuefeng Lv ◽  
Fenglei Niu

Passive containment cooling system is an important safety-related system in AP1000 nuclear power plant, by which heat produced in reactor is transferred to the heat sink–atmosphere based on natural circulation. So it is important to the plant safety whether the system can work well or not in the seismic hazard. Since the system operation is independent of human interfere or the operation of outside equipments, the reliability of the system is improved, however, physical process failure become one of the important contributors to the system operation failure because natural circulation may not keep when the system configuration is different from the design. So it is necessary to analyze the system reliability in seismic situation. The equipment failure probability under earthquake is a function of the peak ground acceleration which is stochastic, and the fault tree method used in traditional probability safety assessment (PSA) for system reliability analysis is not power enough to deal with conditional probability. In this paper, a new analysis method for system reliability evaluate at seismic situation based on Monte Carlo (MC) simulation is put forward, and annual failure probability of passive containment cooling system in AP1000 in seismic hazard is calculated, the result is according with the AP1000 seismic margin evaluation.


Author(s):  
Sungyeol Choi ◽  
Il Soon Hwang ◽  
Jae Hyun Cho ◽  
Chun Bo Shim

Since 1994, Seoul National University (SNU) has developed an innovative future nuclear power based on LBE cooling advanced Partitioning and Transmutation (P&T) approach that leaves no high-level waste (HLW) behind with transmutation reactor named as Proliferation-resistant, Environment-friendly, Accident-tolerant, Continual, and Economical Reactor (PEACER). A small modular lead-bismuth cooled reactor has been designated as Ubiquitous, Robust, Accident-forgiving, Nonproliferating and Ultra-lasting Sustainer (URANUS-40) with a nominal electric power rating of 40 MW (100 MW thermal) that is well suited to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. URANUS-40 is a pool type fast reactor with and an array of heterogeneous hexagonal core, fueled by proven low-enriched uranium dioxide fuels. The primary cooling system is designed to be operated by natural circulation. 3D seismic base isolation system is introduced underneath the entire reactor building allowing an earthquake of 0.5g zero period acceleration (ZPA) for the Safe Shutdown Earthquake (SSE). Also, the proliferation risk can be effectively managed by capsulized core design and a long refueling period (25yr).


Author(s):  
Shengzhi Yu ◽  
Jianjun Wang

On the basis of passive containment cooling system concept, the one-dimensional codes are developed for the analysis of containment thermal-hydraulic characteristics under accident conditions, which can be applied to deal with large dry concrete containment in nuclear power plant. In order to build up the hypothesis flowing path, the containment space is divided into a rising channel and downward ring during the geometric modeling process. In this paper, the physical models, the identification of solving methods and the verification of the codes are introduced. It is assumed that the control volumes take the adiabatic condition in addition to heat exchanger and breaks, and there is no exchange of mass, momentum and energy between each volume in the rising channel and corresponding volume in downward ring. In the analysis, we also assume that the heat in the containment can only be transferred through natural circulation by passive containment cooling system. Furthermore, the break is supposed in the center of bottom of the containment. In this paper, the responses of the containment are predicted with the codes under large LOCA scenario. Under the same conditions, the characteristics of the natural circulation are also analyzed through the codes for the passive containment cooling system. The results can provide some references for the design of the passive containment cooling system.


Algologia ◽  
2021 ◽  
Vol 31 (4) ◽  
pp. 299-319
Author(s):  
V.I. Shcherbak ◽  
◽  
S.I. Genkal ◽  
N.Ye. Semenyuk ◽  
◽  
...  

The paper deals with the long-term dynamics of taxonomic composition of diatom periphyton in the Chornobyl nuclear power plant cooling pond (ChNPP cooling pond) at different stages of its operation: before the accident, after the accident and during the present period. The dominant complex of diatoms was marked by the highest diversity in the period after the accident, due to water temperature decreasing and new habitats appearing. The large-scale water-level drawdown in the present period caused the water table to reduce, and the habitats became less diverse. Owing to this, the number of dominant species decreased. Studying the present-day taxonomic composition of periphytic algae in the ChNPP cooling pond by way of light microscopy and scanning electron microscopy made it possible to identify 141 diatom species, represented by 143 infraspecific taxa, from 45 genera, 20 families, 12 orders and 3 classes. 14 species and infraspecific taxa of diatoms from genera Amphora, Cocconeis, Gomphonema, Hippodonta, Karayevia, Navicula, Placoneis, Planothidium, Psammothidium, Sellaphora are new for Ukrainian flora. High contamination of the ChNPP cooling pond with man-made radionuclides 90Sr, 137Cs and the large-scale water-level drawdown did not cause a significant degradation of diatom periphyton, which, in new ecological conditions, is distinguished by high taxonomic diversity and spatial heterogeneity.


Author(s):  
Gyo¨rgy E´zso¨l ◽  
Ga´bor Baranyai ◽  
La´szlo´ Perneczky ◽  
La´szlo´ Szabados ◽  
Iva´n To´th

Research and development programs of high safety significance have been going on for nuclear power plants of VVER-440/213 type to apply the in-vessel corium retention concept. The in-vessel retention (IVR) concept is based on the external reactor vessel cooling (ERVC) with the main objective to prove that the reactor pressure vessel (RPV) integrity can be preserved in accident sequences leading to core melt. One of the bases of programs was the SARNET project of European Commission, which focused on confirming the capability of the ASTEC code to simulate IVR, calculating thermal load caused by the corium. The ERVC concept is applied to the Paks nuclear power plant of VVER-440/213 type. For the experimental modelling of the ERVC, the CERES (Cooling Effectiveness on Reactor External Surface) facility was designed and constructed. The facility is a scaled down model of the cooling system intended to apply to the Paks NPP with 1:40 scaling ratio for the vessel external surface and 1:1 for the elevations giving the driving force for natural circulation. The heat load supplied to the model is provided by electric heaters. A large number of temperature, pressure, level, void and flow measurements are installed. A RELAP model of the CERES facility was developed and tested by pre-test results.


2015 ◽  
Vol 751 ◽  
pp. 263-267
Author(s):  
Su'ud Zaki

In post Fukushima nuclear accidents inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Gas cooled fast reactors is one of the important candidate of 4th generation nuclear power plant and in this paper the safety analysis related to unprotected loss of flow in small long life gas cooled fast reactors has been performed. Accident analysis of unprotected loss of flow include coupled neutronic and thermal hydraulic analysis which include adiabatic model in nodal approach of time dependent multigroup diffusion equations. The thermal hydraulic model include transient model in the core, steam generator, and related systems. Natural circulation based heat removal system is important to ensure inherent safety capability during unprotected accidents. Therefore the system similar to RVACS (reactor vessel auxiliary cooling system) is investigated. As the results some simulations for small 60 MWt gas cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations.


Author(s):  
Cheol Woo Kim ◽  
Seok Jeong Park ◽  
Chul Jin Choi ◽  
Jong Tae Seo

For an optimum recovery from a steam generator tube rupture (SGTR) event, the operators are directed to isolate the steam generator (SG) with ruptured tube as early as possible to minimize the radioactive material release. However, the reactor coolant system (RCS) cooldown and depressurization to the shutdown cooling system (SCS) operation conditions using the intact SG only are hard to achieve unless the ruptured SG is properly cooled since the ruptured SG, which is isolated by operator, remains at high temperature even though the RCS has been cooled down. The effects of intentional back flow from the SG secondary side to the RCS through the ruptured U-tube on the the ruptured SG cooldown were evaluated for the pressurized light water reactor, especially for the Korean standard nuclear power plant (KSNP). In order to evaluate the back flow effect, a series of analyses was conducted using the RELAP5/MOD3 computer code. For the first stage of the analysis, the cooldown process by natural circulation in the SG secondary side was simulated for the initial conditions of the ruptured SG cooldown. In the next analysis stage, two methods of the ruptured SG cooldown by using back flow after RCS cooldown were evaluated. The one is a tube uncovery, which utilizes the steam condensation on the uncovered U-tube surface, and the other is a SG drain and fill. In the former method, SG tubes are exposed to the steam space by draining SG secondary water into the RCS in order to condense the steam directly onto the uncovered tubes. This method showed that the steam condensation decreased SG secondary pressure and temperature rapidly, demonstrating its effectiveness for cooling. However, this process has a limited applicability in cases of that the rupture is located at the lower region. The latter method, draining by back flow and filling using the feedwater system was also found to be effective in ruptured SG cooldown and depressurization even if the rupture occurred at the top of the U-tube. However, the actuation of the feedwater system is a burden to operator since the makeup of cold feedwater was required to complete cooldown by one cycle of the draining and filling. It is concluded that utilization of the intentional back flow from the SG secondary side to the RCS is very effective for rapid cooling of the RCS to the SCS entry conditions.


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