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Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barr...
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Citations
... During an accident in a water reactor, the "blowdown" phase refers to the initial discharge, with a high mass flow rate of hightemperature pressurized coolant from the reactor cooling system into the containment. The intensity of the release is due to the high pressure difference between the cooling system and the containment atmosphere (Noori-Kalkhoran et al., 2016). Two main role-playing factors in such accidents include Blowdown source and as its result, Thermal-Hydraulic distribution inside containment. ...
... Fernández-cosials et al. (2017) provided the overall peak temperature and pressure of the containment of an AP1000 reactor with a detailed three-dimensional representation of the geometry of the whole building by GOTHIC. Containment pressure distribution has been studied in a VVER-1000 reactor using different methods by (Noori-Kalkhoran et al, 2014a, 2014b, 2016. They have applied single-and multi-cell models and CONTAIN 2.0 for simulation of TH parameters inside VVER-1000 containment. ...
... A modified 30-cells model (in comparison to previous 23-cells studies; Noori-Kalkhoran et al., 2016;2014b;2014a) has been applied in this research to simulate the BNPP-1 containment pressurization as result of LBLOCA. This modified version can cover the spot points of TH parameters and hydrogen distribution (not included in this research) in a more detailed and scientific manner and can help siting of ESFs in more efficient coordinates. ...
Nuclear power plants containment plays an important role as last-defined barrier in defense in depth approach against the release of radioactive material to the environment. In this study, a parallel processing couple has been developed to full scope analysis of blowdown source and containment pressurization parameters in a LBLOCA accident. To achieve this goal, primary and secondary loops of a VVER-1000/V446 were first simulated in TRACE V5.0 and steady-state results have been validated against reference data. The second step deals with containment simulation in CONTAIN 2.0 with new modified 30-cells models. A parallel processing interface was developed in MATLAB to couple TRACE and CONTAIN in the break point. Containment average pressure has been fed back to TRACE as forcing function of blowdown source in each time step during pressurization phase (coupling point). Finally, results of blowdown and containment pressurization have been validated against final safety analysis report (FSAR). Results of simulation confirm that the maximum containment pressure can reach 0.36 MPa and 0.395 MPa for this study and FSAR respectively that are lower than the maximum design absolute pressure of 0.46 MPa, so containment maintains its integrity during this accident. Temperature profiles of different control volumes inside containment during accident follow the FSAR profiles in terms of shape and value that show the ability of developed parallel coupling to full scope simulation of accidents accurately.
... Depressurization of primary system corresponds to pressurization of containment, full pressure type in PWR, see e.g. OCED/NEA/ CSNL, 1999, andNoori-Kalkhoran et al., 2016. Following pipe whip, jet impingement, thrust on RPV supports and possible missiles during the first LOCA instants, pressure and temperature loads in containment are created by the discharging two-phase flow. ...
The paper deals with the safety evaluation and the embedded licensing process of the Atucha-II 800 Mwe Pressurized Heavy Water Reactor (PHWR) in Argentina. Atucha-II was designed by Kraftwerk Union (KWU) during the 60’s, the related construction was started/stopped in the 90’s and restarted on 2006, and was connected to the electrical grid in 2014. Because of market policies, the KWU designer could not be directly involved in the licensing process during the first decade of the 2000 millennium: a licensing suited safety evaluation was performed by the Nucleoeléctrica Argentina (NA-SA), utility owner of the Atucha-II, with support of external experts group. A designer-independent assessment was performed having access to the installed systems and components other than the relevant design documents. Large core size (related to Pressurized Water Reactor – PWR – producing the same thermal power), presence in the core of natural uranium and heavy water fluids, i.e. the coolant and the moderator driven by two circulation loops with different average temperatures, characterize the system design. In those conditions, during the early phase of a depressurization transient, the ‘hot’ coolant vaporizes and the colder moderator remains in the liquid phase: a positive void coefficient is created. The relevance of the Large Break Loss of Coolant Accident (LBLOCA) in safety assessment is discussed with emphasis given to the system design features, the approach pursued in the analysis and the key results. Break opening time and time of occurrence of the safety boron injection affect, during the early period of the transient, the propagation of (negative) pressure wave, the fluid flashing, the heat transfer and the neutron flux: a fission power excursion is expected to occur. The analysis of a complex three-dimensional situation, considering the unavoidable uncertainties associated with the computation, demonstrates that safety limits are preserved.
... In some of these studies, containment parameters due to LB-LOCA were simulated by using different tools and models. Noori-kalkhoran et al. have applied different tools for the simulation of thermal-hydraulic parameters of containment due to DECL: CONTAIN code, single-cell, and multi-cell models [5,6]. The GOTHIC code has been used widely to simulate the parameters in IRIS [7], ABWR [8], BWR Mark III [9], and PWR [10] containments. ...
In this study, thermal–hydraulic parameters inside the containment of a WWER-1000/v446 nuclear power plant are simulated in a double-ended cold leg accident for short and long times (by using CONTAIN 2.0 and MELCOR 1.8.6 codes), and the effect of the spray system as an engineering safety feature on parameters mitigation is analyzed with the former code. Along with the development of the accident from design basis accident to beyond design basis accident, the Zircaloy–steam reaction becomes the source of in-vessel hydrogen generation. Hydrogen distribution inside the containment is simulated for a long time (using CONTAIN and MELCOR), and the effect of recombiners on its mitigation is analyzed (using MELCOR). Thermal–hydraulic parameters and hydrogen distribution profiles are presented as the outcome of the investigation. By activating the spray system, the peak points of pressure and temperature occur in the short time and remain below the maximum design values along the accident time. It is also shown that recombiners have a reliable effect on reducing the hydrogen concentration below flame propagation limit in the accident localization area. The parameters predicted by CONTAIN and MELCOR are in good agreement with the final safety analysis report. The noted discrepancies are discussed and explained.
One of the main goals of severe accident management strategies is to maintain containment integrity to prevent radioactive release into the environment. As a result, containment internal loads need to be investigated during a postulated accident. In the present study, short-term containment thermal-hydraulic (TH) parameters of a VVER-1000/V446 reactor are analysed during a LB-LOCA. To achieve this goal, in the first step as-built 3D structure of VVER-1000/V446 containment was modelled in detail by using AutoCAD. The AutoCAD model has been processed to be prepared for GOTHIC 3D input. Meanwhile, an equivalent GOTHIC lumped parameter (LP) model is also prepared to validate the modelling procedure and results against the FSAR. Finally, LP profiles and 3D TH contour results of GOTHIC code were presented and discussed. LP results show a close agreement with FSAR reference and can approve the accuracy of the simulation procedure. 3D contours present all-coordinate detailed TH parameters versus time inside the containment. Spatial distribution of TH parameters and minor short-term effects of spray as Engineering Safety Features (ESFs) to tackle the containment pressurization can be evaluated by employing these contours. 3D simulation results can provide advantages for the precise locating and installation of ESFs in design and operation to achieve the highest efficiency in case of containment accident.
Containment safety analyses are usually performed applying conservative assumptions. These conservatisms are taken to ensure the safety margins overcoming the lack of knowledge in physical phenomena involved. With these assumptions, the results are unrealistically conservative, but also satisfy the regulatory authority requirements. However, in order to obtain an optimal design and reactor operation, conservatism may be limited from the safety analyses. Consequently, in 1989, the U.S. NRC modified the licensing requirements allowing the use of realistic methods if uncertainties are identified and quantified.
Nevertheless, the containment building, and the in-containment equipment are still licensed based on the pressure and temperature obtained with conservative containment calculations under the lumped parameters framework. The average containment pressure calculated with the lumped parameters approach is fairly representative of the containment pressure, as the pressurization develops quite homogeneous. On contrary, the containment temperature calculated by the lumped parameters approach is an averaged temperature and does not necessarily represent its heterogeneous nature.
Therefore, more realistic containment analyses accounting for local conditions are needed, and consequently, a modeling guideline for high-detailed evaluation models with the GOTHIC code is proposed. It is based in three main steps: development of a 3D detailed Computer-Aided Design (CAD) model; adapting the detailed CAD model to obtain a simplified version of the geometry; making use of the geometric data from the simplified CAD model, a three-dimensional thermal-hydraulic evaluation model is conformed.
In addition, when a complex system (e.g. a containment building) is modeled with a CFD code like GOTHIC, it is important to assure that results are not dependent on the mesh chosen. That means that results will not change substantially even if the mesh is subsequently refined. This process is clear when the employed code separates the fluid region from the solids (as traditional CFD codes do) but becomes more complicated when the computational cells includes fluid and solids, as the case of the porous media approach.
According to that, a mesh sensitivity study using 12 different meshes was performed to analyze the mesh independence in the 3D GOTHIC containment evaluation model under the porous media framework. When the traditional CFD method of mesh refinement is applied, results become not conclusive. There are no big discrepancies when key parameters are compared (averaged temperature and pressure peaks). Nevertheless, it was found that cell aspect ratio influences negatively in the results of highly blocked cells generating numerical instabilities during the calculation. In addition, size differences also influence the fluid velocity field modifying the flow patterns, and therefore, the local temperature profiles. Considering the lessons learned from this study, different recommendations are suggested to be applied in case of performing a mesh independence studies using porous CFDs codes like GOTHIC.
To test the developed 3D containment evaluation model, an application case was performed for assessing the generic equipment qualification criteria when local data for pressure and temperature is obtained. The transient simulated was a DEGB-LOCA located at the cold leg. Results showed how the temperature heterogeneity in the containment compartments makes inadequate averaged values like those obtained with evaluation models under the lumped parameters framework.
Since a single Best-Estimate (BE) calculation brings results with unknown accuracy, an uncertainty analysis is required to estimate the solution precision. Nonetheless, BEPU analyses has been historically applied to RCS transient analysis, but it starts to be also applied to containment analysis with limited scope. Consequently, accounting for the experiences that many analyst and researchers have gathered along the last three decades, a BEPU methodology, with the containment safety analysis in mind, is proposed. Conservative assumptions are avoided by developing best estimate containment evaluation models. Uncertainties are quantified and propagated through the code in order to obtain the results accuracy. The proposed methodology is defined in a hierarchical structure, similar to the U.S. NRC Regulatory Guide 1203. It is divided in two main blocks; the first is related to the best estimate model setting; and the second to the uncertainty treatment.
The BEPU-CSA methodology was applied for the same analysis performed for the equipment qualification criteria above named. It was started with a “traditional” BEPU analysis based on the non-parametric tolerance interval calculation applying the famous Wilks approach with no segregation between epistemic and random uncertainties. Two series were calculated, one with a sample set of 59 elements for a one-sided tolerance region (Wilks-OS), and other with a sample set of 93 elements for a two-sided region (Wilks-TS).
Then, a method based on the LHS for obtaining similar bounds as in the case of applying the Wilks formula is also discussed, but with a reduced number of cases. Two series were also calculated, one with a sample set of 20 elements for a one-sided tolerance region (LHS-20), and other with a sample set of 40 elements for a two-sided region (LHS-40).
For both, Wilks and LHS series, maximum values obtained for pressure and temperature are quite similar, resulting more conservative the LHS sampling even having a smaller sample size.
In addition, a second order uncertainty analysis, where the epistemic uncertainties were treated based on the Dempster-Shafer theory with LHS sampling and the random uncertainties by applying the Wilks method, was compared against these obtained with the Wilks and LHS methods. A set of 30 Wilks series were obtained, being this the result of the LHS sample with size 30 for the outer loop, and 59 runs for every of the 30 LHS sample elements, which led to a 1770 code runs.
Analogously, another second order uncertainty analysis was also performed based entirely on the LHS sampling method (denominated 2nd Order LHS-20), were the epistemic uncertainty sample size was set to 20, and the random uncertainty sample size also to 20 based on the results obtained in the comparison between the Wilks and LHS series above commented.
Both methods, the LHS-Wilks and the 2nd order LHS-20, showed similar results between them, but it was observed that, when the uncertainties are segregated between random and epistemic, pressure and temperature bounds become larger than that obtained in the “traditional” BEPU analyses. This is caused by the underestimation when uniformity is assumed over imprecise uncertainties that are in fact governed by a random nature, as is the case of the initial conditions. However, when uncertainties are purely epistemic, and it is mean with purely epistemic as a property of the analyst and not over the parameter itself, uniformity resulted adequate.
At the end, a sensitivity analysis was performed over the LHS-Wilks calculation showing that for short-term analysis, only a few of the parameters analyzed resulted correlated with the maximum pressure and temperature obtained. In addition, it was observed that uncertainties affecting the averaged values differs from that affecting local values, indicating that dominant phenomena may differs at different scales, something to be accounted in scaling analyses.
The paper deals with the safety evaluation and the embedded licensing process of the Atucha-II 800 Mwe Pressurized Heavy Water Reactor (PHWR) in Argentina. Atucha-II was designed by Kraftwerk Union (KWU) during the 60’s, the related construction was
started/stopped in the 90’s and restarted on 2006, and was connected to the electrical grid in 2014. Because of market policies, the KWU designer could not be directly involved
in the licensing process during the first decade of the 2000 millennium: a licensing suited safety evaluation was performed by the Nucleoeléctrica Argentina (NA-SA), utility owner of the Atucha-II, with support of external experts group. A designer-independent assessment was performed having access to the installed systems and
components other than the relevant design documents. Large core size (related to Pressurized Water Reactor – PWR – producing the same thermal power), presence in the core of natural uranium and heavy water fluids, i.e. the coolant and the moderator driven by two circulation loops with different average temperatures, characterize
the system design. In those conditions, during the early phase of a depressurization transient, the ‘hot’ coolant vaporizes and the colder moderator remains in the liquid phase: a positive void coefficient is created. The relevance of the Large Break Loss of Coolant Accident (LBLOCA) in safety assessment is discussed with emphasis given to the system design features, the approach pursued in the analysis and the key results. Break opening time and time of occurrence of the safety boron injection affect, during the early period of the transient, the propagation of (negative) pressure wave, the fluid flashing, the heat transfer and the neutron flux: a fission power excursion is expected to occur. The analysis of a complex three-dimensional situation, considering the unavoidable uncertainties associated with the computation, demonstrates that safety limits are preserved.
Regarding the importance of radioactive materials dispersion in the environment and their effect on living organism's well-being, this article examines the atmospheric dispersion of these materials around the Bushehr nuclear power plant (BNPP) in a hypothetical severe accident. One of practical dispersion models is AERMOD which is based on Gaussian distribution. This model which is an extension of Gaussian model is normally used for predicting the gaseous pollution. The aim of this study is to investigate the dispersion of radioactive materials in BNPP's surroundings due to a specific type of Large Break LOCA (LBLOCA) accident. In this study, the wind direction, wind speed, and the topological conditions of the power plant's environment are considered. In order to execute this model, the meteorology and topological data of the region have been provided. Besides, using the CONTAIN code, the data about released radioactive materials from the containment have been provided. The double ended cold leg break (DECL) accident in containment has been modeled and its validity has been checked with FSAR. Extracting all required data, the simulation is done in AERMOD and the volumetric concentration and activity of fission products are calculated. The volumetric activity of fission products in Tangestan, Bandar-E-Bushehr and Delvar cities are evaluated as 7.5658E+07, 7.5136E+07 and 2.9655E+06 (Bq/m³), respectively. From results it could be seen that the most dispersion of the radioactive materials are in direction of the non-residential regions, although the pollution of the cities around the power plant is inevitable.
