• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.623-639
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• 2023

The specific configuration of the safety systems in VVER-1000/V320 reactors allows a comprehensive study of the Loss of Coolant Accident (LOCA). In the present paper, a verification of the success criteria of the event trees headers for the medium and large break LOCA sequences is conducted. A detailed TRACEV5P5 thermal-hydraulic model of the reactor has been developed, including all safety systems. When analyzing the results of all sequences, some conservatism is observed in certain specific configurations as the success criterion of some headers is not consistent with the classic PSA level 1. Therefore, new proposals for the LOCA event trees are performed based on a reconfiguration of LOCA break ranges and the use of the expanded event trees approach.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2830-2838
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• 2021

This work deals with the assessment of the burnup capabilities of the Serpent Monte Carlo code to predict spent nuclear fuel (SNF) isotopic concentrations for low-enriched uranium (LEU) fuel at different burnup levels up to 47 MWd/kgU. The irradiation of six UO₂ experimental samples in three different VVER-1000 reactor units has been simulated and the predicted concentrations of actinides up to ²⁴⁴Cm have been compared with the corresponding measured values. The results show a global good agreement between calculated and experimental concentrations, in several cases within the margins of the nuclear data uncertainties and in a few cases even within the reported experimental uncertainties. The differences in the performances of the JEFF3.1.1, ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries (NDLs) have also been assessed and the use of the newly released ENDF/B-VIII.0 library has shown an increased accuracy in the prediction of the C/E's for some of the actinides considered, particularly for the plutonium isotopes. This work represents a step forward towards the validation of advanced simulation tools against post irradiation experimental data and the obtained results provide an evidence of the capabilities of the Serpent Monte-Carlo code with the associated modern NDLs to accurately compute SNF nuclide inventory concentrations for VVER-1000 type reactors.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.1-18
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• 2021

The criticality analysis of VVER-1000 mock-up benchmark experiments from the LR-0 research reactor operated by the Research Center Rez in the Czech Republic has been conducted with the MCS Monte Carlo code developed at the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The main purpose of this work is to evaluate the newest ENDF/B-VIII.0 nuclear data library against the VVER-1000 mock-up integral experiments and to validate the criticality analysis capability of MCS for light water reactors with hexagonal fuel lattices. A preliminary code/code comparison between MCS and MCNP6 is first conducted to verify the suitability of MCS for the benchmark interpretation, then the validation against experimental data is performed with both ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. The investigated experimental data comprises six experimental critical configurations and four experimental pin-by-pin power maps. The MCS and MCNP6 inputs used for the criticality analysis of the VVER-1000 mock-up are available as supplementary material of this article.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1603-1610
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• 2020

Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.561-574
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• 2009

First ex-vessel core catcher has been applied to the practical design of NPPs with VVER-1000 reactors built in China (Tyanvan) and India (Kudankulam) for severe accident management (SAM) and mitigation of SA consequences. The paper presents the concept and basic design of this crucible-type core catcher as well as an evaluation of its efficiency. The important role of oxidic sacrificial material is discussed. Insight into the behaviour of the molten pool, which forms in the catcher after core relocation from the reactor vessel, is provided. It is shown that heat loads on the water-cooled vessel walls are kept within acceptable limits and that the necessary margins for departure from nucleate boiling (DNB) and of vessel failure caused by thermo-mechanical stress are satisfactorily provided for.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.125-137
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• 2019

The purpose of this paper is to study the source term behavior after severe accidents by using a semi-kinetic model for simulation and calculation of in-containment activity. The reactor containment specification and the safety features of the containment under different accident conditions play a great role in evaluating the in-containment activity. Assuming in-vessel and instantaneous release of radioactivity into the containment, the behavior of in-containment isotopic activity is studied for noble gasses (Kr and Xe) and the more volatile elements of iodine, cesium, and aerosols such as Te, Rb and Sr as illustrative examples of source term release under LOCA conditions. The results of the activity removal mechanisms indicates that the impact of volumetric leakage rate for noble gasses is important during the accident, while the influence of deposition on the containment surfaces for cesium, mainly iodine isotopes and aerosol has the largest contribution in removal of activity during evolution of the accident.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1176-1179
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• 2019

Measurements of activity concentrations of radionuclides in atmospheric releases were performed in 2017-2018 at vent stacks of seven Russian nuclear power plants. The selected instruments and research methods, with detection limits significantly lower than the existing detection limit of Russian NPPs routine control, allowed to reliably determine up to 26 radionuclides. Analysis of experimental data allows to determine the list of radionuclides for calculation the effective dose rates to public and the permissible annual discharge levels for each Russian NPP. Radiocarbon is determined as major contributor for the dose from the atmospheric releases of LWGR reactors - up to 98% for EGP-6 and RBMK-1000 (Smolensk NPP) reactors. For PWR reactors (VVER) radionuclides contribution to the annual dose from atmospheric releases is more complicated, but, in general, dose is formed by tritium, $^{14}C$ and noble gases. The special monitoring results with ranking of measured radionuclides according to their contribution to the effective dose makes it possible to optimize the list of controlled radionuclides in airborne releases of Russian NPPs from 94 to 8-16 for different NPPs.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2928-2938
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• 2020

A light water nuclear Reactor has been exergy analyzed, and the rate of irreversible exergy loss and exergy destruction is calculated for each of its components. The ratio of these losses compared to the total input exergy loss is calculated, which shows that most irreversible losses occur in the reactors, turbines, steam generators, respectively, as well as in the downstream operations. The main aim of this paper is to optimize the power plant using an innovative firefly algorithm and then to propose a novel strategy to improve the overall performance of the plant. As shown in the results, the exergy destruction rate of the plant decreased by 1.18% using the firefly method, and the exergy efficiency of the plant reached 29.3% comparing to the operational amount of 28.99%. Also, the results of the firefly optimization process compared to the Genetic algorithm and gravitational search algorithm to study the accuracy of the model for exergy analysis fitness problems in the power plants and the results of this comparison has shown that the results are nearly similar in the mentioned methods. However, the firefly is faster and more accurate in limited iterations.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.169-174
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• 2016

This article describes a comprehensive methodology for the evaluation of the middle part of nuclear fuel cycles. Evaluation of fuel cycles is basically divided into two parts. The first comprises nuclear calculation, i.e., creation of the strategy for nuclear fuel reloading and core design calculations. The second part is the business-economic evaluation of the selected reloading strategy, which can be done either by financial analysis or economic analysis. The financial analysis incorporates the perspectives of a company while the economic analysis can be used primarily by national economists or politicians. This methodology was applied to a case study that is focused on impacts of switching from a 12-month to an 18-month fuel cycle strategy for Water-Water Energetic Reactor (VVER)-1000 reactors.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1213-1221
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• 2020

Chromium-nickel alloy 42XNM (XHM-1, Bochvalloy) is considered as a promising material for future generations of nuclear reactors, primarily as a material for the fuel elements shells in the development of accident tolerant fuel. However, as with most nickel-based alloys, 42ХNМ is characterized by a sharp decrease in plastic properties in the temperature range of (500-900)℃. This effect is enhanced by neutron irradiation. Preliminary tests of ring samples of 42XNM alloy (after irradiation as a part of the VVER-1000 control system) in the temperature range of ductility failure showed that the standard technique for processing tensile diagrams does not allow to evaluate the plastic properties correctly at low strains. Therefore, in this work, the alternative method for testing ring samples from materials with low plastic characteristics was developed. It was shown that the minimum value of the permanent strain of the irradiated 42XNM alloy in the temperature range of (500-1100)℃, determined by the alternative method, was ~1.6% at 750 ℃.