VVER-1000 Nuclear Power Plant Research Articles

Influence of mixed core in the radionuclides releases and hydrogen production for VVER-1000

This paper aims to compare the behaviour of VVER-1000 Nuclear Power Plants (NPPs) under DEC-A/B conditions, contemplating different fuel assemblies considered in mixed cores. The main characteristics of both fuel assemblies are described in this paper, with a focus on highlighting the key differences in mass composition.There are three models under consideration: Model A, Model B and Model C. Model A consists of a core entirely filled with fuel assemblies equipped with grids made of Zr-1 %Nb, referred to as “Type A” through the paper. Model B comprises a core filled with fuel assemblies that have grids made of Alloy 718 (aka Inconel), referred to “Type B”. Lastly, Model C features a mixed core of 151 fuel assemblies with Zr-1 %Nb grids and 12 fuel assemblies with Inconel grids.Initially, the primary distinctions between these fuel assemblies types are the number and composition of their spacer grids. Type A fuel assemblies have 13 grids made of Zirconium, while Type B fuel assemblies have 16 grids made of Inconel. It was initially anticipated that the reduction in the mass of zirconium available for oxidation would lead to lower hydrogen production in Model B compared to Model A. However, it was observed that the Inconel grids in Type B fuel assemblies provide enhanced structural support, prolonging the oxidation reaction and yielding more hydrogen during the initial oxidation phase compared to Type A fuel assemblies.Moving on to the mixed core configuration (Model C), the Inconel grids in Type B fuel assembly also offer superior structural support. This results in an extended oxidation process. The Type A fuel assemblies having a significant zirconium mass, oxidation in intensified, causing a rise in temperature due the exothermic reaction and leading to the producing of a greater amount of hydrogen compared to the other two models. This pattern is also observed in radionuclide releases.In conclusion, this study sheds light on the behaviour of VVER-1000 NPPs under DEC-A/B conditions, considering different fuel assembly types and mixed core configurations. The results indicate that the choice of grid materials significantly influences the oxidation process and hydrogen production, with Inconel grids exhibiting extended oxidation and hydrogen generation behaviour in both pure and mixed core setups.

Assessment of spray system capabilities in reduction of containment pressure and deposition of fission products during a severe accident in VVER-1000

In this study, we have investigated the spray system's capabilities to reduce fission products in the containment atmosphere and control the containment pressure in a VVER-1000 nuclear power plant. Large Break Loss of Coolant Accident (LB-LOCA) without access to the active part of the Emergency Core Cooling System (ECCS), as a worst-case severe accident, has been investigated as a case study. Only containment is simulated via the MELCOR code, while leakage of various materials from the nuclear power plant's core and the primary circuit is considered as the boundary conditions. Following the Fukushima accident in 2011, many nuclear safety regulatory organizations focused on using portable equipment that is essential for decay heat removal in nuclear power plants during severe accidents. Based on the literature studies, the accident management scenarios have changed recently since severe accidents inside nuclear power plants have worried many scientists and the public. Evaluating the current spray system in a severe condition based on which this system was not designed is vital and novel. We considered that the spray system would be fully restored and perform its safety function using the existing portable equipment to depressurize the containment and deposition of the fission products. To calculate the maximum allowed delay time of the spray system recovery, the accident was modeled without the operation of the spray system while a safety criterion was considered to avoid reaching the ultimate pressure of containment. For a more detailed evaluation of the spray system, two different operating modes including the activation of one spray train and the activation of both spray trains have been selected. According to the obtained results, if the spray system is revived before 4,000s from the beginning of the accident, this system would be able to reasonably mitigate the accident consequences. Moreover, without the need for any external water supply source for the spray system, this system can be permanently fed from the containment sump for at least 800,000s after the initiation of the accident, requiring the relevant heat exchanger of the spray system to be recovered for this period.

Порівняльний метод кваліфікації систем безпеки ядерних енергоустановок з ВВЕР‑1000 та АР1000

A deterministic method of analyzing conditions for ensuring the safe operation of VVER-1000 and AP1000 nuclear power plants has been developed. The method is based on the comparison of the provision of the required safety functions by critical (minimum permissible) configurations of safety systems for accident management. During the analysis, the composition and information on the technical characteristics of the safety systems of the VVER-1000/320 and AP1000 reactor installations, which are designed to ensure the safety functions of feeding and cooling the reactor/PG, were considered. On the basis of the developed method, the relative criteria and safety conditions (for example — the criterion of safety — the maximum temperature of the heating element shell; the condition of safety is not to exceed the maximum design limit of damage to the twelfth (the temperature of the twelfth shell is 1,200 °C) for nuclear power plants with VVER-1000 and АР1000 reactors in terms of safety functions for emergency fueling and cooling of the reactor core and steam generators are determined. Preliminary estimates of the relative criteria and conditions for ensuring nuclear safety during accident management of natural circulation circuits of passive safety systems AP1000 obtained taking into account data on their composition and technical characteristics, the table showed that the required height of natural circulation circuits of passive safety systems should be several hundred meters. Based on the established relative criteria and safety conditions, the need for additional modernization of the АР1000 safety systems to successfully ensure the safety functions of emergency coolant/feedwater supply and core and steam generator cooling is preliminarily determined.

Radiological Safety Analysis for a Hypothetical Accident of a Generic VVER-1000 Nuclear Power Plant

Atmospheric dispersion modelling and radiological safety analysis have been performed for a postulated accident scenario of a generic VVER-1000 nuclear power plant using the HotSpot Health Physics code. The total effective dose equivalent (TEDE), the respiratory time-integrated air concentration, and the ground deposition concentration are calculated considering site-specific meteorological conditions. The results show that the maximum TEDE and ground deposition concentration values of 3.69E – 01 Sv and 3.80E + 06 kBq/m2 occurred at downwind distance of 0.18 km from the release point. This maximum TEDE value is recorded within a distance where public occupation is restricted. The TEDE values at distances of 5.0 km and beyond where public occupation is likely to be found are far below the annual regulatory limits of 1 mSv from public exposure in a year even in the event of worse accident scenario as set in IAEA Safety Standard No. GSR Part 3; no action related specifically to the public exposure is required. The released radionuclides might be transported to long distances but will not have any harmful effect on the public. The direction of the radionuclide emission from the release point is towards the north east. It is observed that the organ with the highest value of committed effective dose equivalent (CEDE) appears to be the thyroid. It was followed by the bone surface, lung, red marrow, and lower large intestine wall in order of decreasing CEDE value. Radionuclides including I-131, I-133, Sr-89, Cs-134, Ba-140, Xe-133, and Xe-135 were found to be the main contributors to the CEDE.

NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE

Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) of the European Commission (EC) in Holland (Europe) for a Pressurized Water Reactor (PWR). The obtained results of both projects have shown very large discrepancies between the used severe accident codes for both reactor types BWR and PWR. Consequently, the results for a real plant analysis by these integral codes, may not be correct after the beginning of core melt. Discrepancies of results of ex-vessel phenomena in the containment between the codes are in general larger. Therefore, there is a strong need for a reliable new generation mechanistic severe accident code which can simulate severe accident scenarios from an initiating event till containment failure with better accuracy not only for existing light water reactors but also for new generation IV reactor types. SAMPSON mechanistic ex-vessel modules coupled with SCDAPSIM and a new thermal-hydraulic module ASYST-ISA with particularly newly developed options for the reactor coolant system (RCS) and material properties applicable to new reactor deigns, is proposed as a best etimate new generation severe accident code for several reasons which are described in this paper.Keywords: Severe accident, SAMPSON, SCDAPSIM, ASYST-ISA, Steam explosion, Hydrogen detonation

Human Error Probability Quantification using SPAR-H Method: Total Loss of Feedwater case study for VVER-1000

The primary side Feed & Bleed (F&B) recovery strategy is the last device to avoid core damage in the event of loss of Ultimate Heat Sink (UHS) accident. In the context of accident management, besides multiple safety systems’ well-functioning, human factor has a significant role in avoiding the extreme scenarios. In this research, it is tried to quantify Human Error Probability (HEP) of operators’ interference to initiate F&B recovery procedure when the Total Loss of Feedwater (TLFW) accident occurs in a typical VVER-1000 Nuclear Power Plant (NPP) via Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) Method. Among those factors that affect human performance is the available time for the operators’ late interference in order to reestablish natural circulation to remove the residual heat after the reactor scrammed which is computed using the best estimate code RELAP5 Mod 3.2. It is revealed that the diagnosis and the action HEPs of the operators are highly dependent and the combined HEP value is too risky with an unacceptable amount of uncertainty. It is concluded that the accident management procedure must be redesigned so as to eliminate diagnosis and action dependency.

Improving the Efficiency of a Nuclear Power Plant Using a Thermoelectric Cogeneration System

The efficiencies of nuclear power plants are rather poor having the ratio %30 by using the conventional energy/exergy tools. According to that information, large amount of energy is wasted during condensation and thrown out to the environment. Thermoelectric generator (TEG) system has a potential to be used as a heat exchanging technology to produce power with a relatively low efficiency (about 5%) and it can transform the temperature difference into electricity and generate clean electrical energy. In the present study, we offer a novel system to recover the waste heat from a VVER-1000 nuclear power plant. The heat transfer of the TEG is analyzed numerically with respect to the various temperature ranges and constant mass flow rate of the exhaust steam entering the system. In the analyses, different hot temperature ranges (35ºC, 45ºC and 55ºC) and a constant cold temperature (i.e. 18ºC) are used for a HZ-20 thermoelectric module and it has been proven that the designed TEG can produce the maximum output power of 76,956 MW for a temperature difference ∆T=37 and the conversion efficiency of 3.854% sits. The TEG is designed for the condenser of a 1000 MW nuclear power plant. It's shown that about 2.0% increasing in the power plant efficiency is expected by using the selected thermoelectric generator in the condensation cycle.Article History: Received: July 15th 2017; Received: October 17th 2017; Accepted: February 13rd 2018; Available onlineHow to Cite This Article: Terzi, R. and Kurt, E. (2018), Improving the efficiency of a nuclear power plant using a thermoelectric cogeneration system, Int. Journal of Renewable Energy Development, 7(1), 77-84.https://doi.org/10.14710/ijred.7.1.77-84

Cancer risk incidence from hypothetical accident of VVER-1000 nuclear power plant based on BEIR VII model

AbstractBackgroundSafety is a mandatory issue during the operation of a nuclear power plant. A nuclear reactor can have some atmospheric dispersion due to any errors in the safety system.PurposeThe aim of this study is to estimate the cancer risk incidence for different body organs due to accidentally released radionuclides from Bushehr Nuclear Power Plant (BNPP).Materials and methodsThe assumed hypothesis was atmospheric dispersion of radionuclide into the environment due to the safety failure of BNPP. Total effective dose equivalent (TEDE) from radionuclide diffusion in the medium was calculated using HOTSPOT code at two different wind speeds. Finally, the risk of cancer incidence for different organs of male and female sex has been estimated by Biologic Effects of Ionizing Radiation (BEIR) VII model.ResultsThe results showed that with increasing the exposure age and attained age, the risk of cancer incidence for different organs is decreased. The value of TEDE was increased at lower wind speed. The most probable organ for cancer incidence at different levels of TEDE in male and female sex was colon and bladder, respectively. On the other hand, prostate and uterus had the lowest radiation sensitivity and cancer risk incidence in male and female sex, respectively. Increasing the wind speed reduces the risk of cancer incidence for all of organs understudy.ConclusionBased on the obtained results, it can be concluded that the younger persons are more subject to the cancer risk incidence because of both the intrinsically greater radio-sensitivity of their organs and their longer remaining life expectancy during which a cancer may develop. The overall risk of cancer incidence as well as the site specific solid cancer incidence were highly dependent to the sex of exposed person, so that the female sex was more exposed to the cancer risk incidence at all of the irradiation levels understudy.

Comparison of Two Control Programs of the VVER-1000Nuclear Power Unit Using Regression Data Mining Models

A load-following mode of nuclear power plants (NPP) is a complicated procedure, since there are significant changes in many interrelated processes. In order to show which control program (CP) of NPP is better to use, data mining (DM) techniques can be introduced. This study proposes a DM approach in order to show a possibility of using DM regression models for NPP. The datasets for DM were obtained by simulating two static CP of VVER-1000 NPP in Simulink software of Matlab program package.

Analyses of Feedwater Trip With SBO Sequence of VVER1000 Reactor

In the Czech Republic, as a follow-up, a consortium of research organizations and universities has decided to simulate selected stress tests’ scenarios, in station blackout (SBO) and the loss of ultimate heat sink (LoUHS), with the aim to verify the national stress report and to analyze time response of respective source term releases. These activities are carried out in the frame of the project prevention, preparedness, and mitigation of consequences of severe accident (SA) at Czech NPPs in relation to lessons learned from stress tests after Fukushima, financed by the Ministry of Interior. The Research Centre Rez has been working on a methods for estimation of leakages and consequences of releases (MELCOR) model for VVER1000 nuclear power plant (NPP) starting with a plant systems nodalization. The aim was to benchmark the MELCOR model with the validated TRAC/RELAP advanced computational engine (TRACE) model, first comparing the steady state and continuing in a long-term SBO plus another event until the beginning of the SA. The presented work is based on the previous paper from the ICONE 23rd Conference hosted in Japan. It focuses mainly on the comparison of the thermohydraulics of the two models created in MELCOR and TRACE codes as outcome of the “Fukushima project.” After that, preliminary general results of the SA progression showing the hydrogen production and the relocation phenomena will be shortly discussed. This scenario is considered closed after some seconds to the break of the lower head. It is important to note that this paper is a substantial update of a previous one (Mazzini et al., 2015, “Analyses of SBO Sequence of VVER1000 Reactor Using TRACE and MELCOR Codes,” 23rd International Conference on Nuclear Engineering (ICONE23)) and although it contains the same descriptive sections, all the results are new.

Two-Phase Flow Field Simulation of Horizontal Steam Generators

Abstract The analysis of steam generators as an interface between primary and secondary circuits in light water nuclear power plants is crucial in terms of safety and design issues. VVER-1000 nuclear power plants use horizontal steam generators which demand a detailed thermal hydraulics investigation in order to predict their behavior during normal and transient operational conditions. Two phase flow field simulation on adjacent tube bundles is important in obtaining logical numerical results. However, the complexity of the tube bundles, due to geometry and arrangement, makes it complicated. Employment of porous media is suggested to simplify numerical modeling. This study presents the use of porous media to simulate the tube bundles within a general-purpose computational fluid dynamics code. Solved governing equations are generalized phase continuity, momentum, and energy equations. Boundary conditions, as one of the main challenges in this numerical analysis, are optimized. The model has been verified and tuned by simple two-dimensional geometry. It is shown that the obtained vapor volume fraction near the cold and hot collectors predict the experimental results more accurately than in previous studies.

Development and qualification of a thermal-hydraulic nodalization for modeling station blackout accident in PSB-VVER test facility

This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.

Safety analysis of LBLOCA in BDBA scenarios for the Bushehr's VVER-1000 nuclear power plant

This paper presents the results of thermal-hydraulic calculations of a large break loss of coolant accident (LBLOCA) analysis for a VVER-1000/V446 unit at Bushehr nuclear power plant (BNPP). LBLOCA is analysis in two different beyond design basis accident (BDBA) scenarios using the RELAP5/MOD3.2 best estimate code. The scenarios are LBLOCA with station blackout (SBO) and LBLOCA with pump re-circulation blockage which have been evaluated in the final safety analysis report (FSAR) of BNPP. A model of VVER-1000 reactor based on Unit 1 of BNPP has been developed for the RELAP5/MOD3.2 thermal-hydraulics code consists of 4-loop primary and secondary systems with all their relevant sub-systems important to safety analysis. The analysis is performed without regard for operator's actions on accident management. The safety analysis is carried out and the results are checked against the acceptance criteria which are the possibility of using water inventory in the emergency core cooling system (ECCS) accumulators and the KWU tanks for core cooling and the available time to operators before the maximum design limit of fuel rod cladding damage is reached. These kinds of analyses are performed to provide the response of monitored plant parameters to identify symptoms available to the operators, timing of the loss of critical safety functions and timing of operator actions to avoid the loss of critical safety functions of core damage. The results of performed analyses show that the operators have 2.9 and 3.1 h for LBLOCA with SBO and LBLOCA with pump re-circulation blockage scenarios, respectively, before the fuel rod cladding rupture. The results are also compared with the BNPP FSAR data.

Analytical validation of operator actions in case of primary to secondary leakage for VVER-1000/V320

This paper presents the results of analytical validation of operator actions in case of “Steam Generator Tube Rupture” (SGTR) for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The purpose of the analyses is to demonstrate the ability to terminate primary to secondary leakage and to indicate an effective strategy for preventing secondary leakage to the environment and in this way to prevent radiological release to the environment. Following depressurization and cooldown of reactor coolant system (RCS) with isolation of the affected steam generator (SG), in these analyses are validated options for post-SGTR cooldown by:•back up filling the ruptured SG;•using letdown system in the affected SG and•by opening Fast Acting Isolation Valve (FAIV) and using Steam Dump Facility to the Condenser (BRU-K).The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs.The RELAP5/MOD3.2 computer code has been used for the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS).This paper is possible through the participation of leading specialists from KNPP.

Steam Line Break investigation at full power reactor for VVER-1000/V320

This paper presents the results of thermal-hydraulic calculation of “Steam Line Break” analysis at full power reactor for VVER-1000/V320 units at Kozloduy Nuclear Power Plant (KNPP), done during the development of symptom based emergency operating procedures (SB EOPs) for this plant. The RELAP5/MOD 3.2 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. A model of VVER-1000 based on Unit 6 of Kozloduy NPP has been developed for the systems thermal-hydraulics code RELAP5/MOD 3.2 at the Institute for Nuclear Research and Nuclear Energy–Bulgarian Academy of Sciences (INRNE–BAS), Sofia.The main purpose of the analysis is to estimate the parameters of the monitored plant which are used to identify symptoms that are used by operators to identify the plant's state and the critical safety function (CSF). The results of the thermal-hydraulic analyses have been used to assist KNPP specialists in analytical validation of EOPs.The performed analysis is based on a previously used bounding approach in analytical validation of SB EOPs. Based on this approach a list of scenarios has been performed, involving a different number of safety systems with or without operator actions. The presented thermal-hydraulic calculations of the accident scenarios involve the loss of CSF “Subcriticality” for VVER-1000/V320 units at KNPP.

ASTECv2/MEDICIS computer code investigation of influence of water content and carbon dioxide content in the concrete on the kinetics of molten corium–concrete interaction

This investigation was done in the frame of SARNET2 European Network of Excellence on Severe Accidents project (6-th Framework Program of the European Commission), joint program activity JPA-2, WP6-4. ASTECv2/MEDICIS computer code was used to simulate the ex-vessel corium–concrete interaction cases. This code, jointly developed by the IRSN (France) and GRS (Germany) for simulation of severe accidents, is now considered as the European reference simulation tool. As a reference nuclear power plant for these ASTECv2/MEDICIS calculations the VVER1000 reactor type was used.The main goal of this paper is to investigate the influence of water content (H2O) and the content of carbon dioxide (CO2) in the concrete on the kinetics of molten corium–concrete interaction (MCCI) between the molten corium and reactor basemat concrete. In this paper, there are investigated different concentrations of H2O and CO2 in the concrete interacting with the corium poured after reactor vessel failure during a severe accident in VVER1000 nuclear power plant.

Neutron and gamma-ray deep penetration calculation through biological concrete shield of VVER-1000 reactor by a new technique based on variance reduction

Reducing the computer run time of neutron and gamma-ray transport calculation as well as obtaining more accurate results are the main objectives of deep penetration problems. In MCNP code, the widely-used variance reduction method, weight window, is applied to reach the above objectives. In the present paper a new method based on the variance reduction technique is introduced. This method is applied to the multilayer shielding structures of serpentine and reinforced concrete which are employed in a VVER-1000 nuclear power plant as biological shields. For applying this method, the thickness and density of shields are reduced to unrealistic values to obtain a suitable estimate for weight windows. Next, the thickness and density are both returned gradually to their real values in iterative MCNP runs until the suitable weight windows for flux and dose rate calculations are obtained. Furthermore, for reducing computer run time, the equivalent fixed surface source is considered instead of critical source. The obtained result of this deep penetration problem is compared with the value which is reported in the Final Safety Analysis Report (FSAR) of this reactor. The comparison shows good consistence between the two outcomes. This consistency satisfies the accuracy of calculation methods.

Study of long-term loss of all AC power supply sources for VVER-1000/V320 in connection with application of new engineering safety features for SAMG

This paper presents the results of analysis for application of a new Severe Accident Management Guideline (SAMG) approach which is specifically applied for VVER-1000/B320 reactor installations. In general, this innovative approach is fully applicable for all the pressurized water reactors from second generation. The purposes of the analysis for the new SAMG approach application are as follows:•To represent suggestions for new engineering safety features application for SAMG strategies.•To assess the applicability of the new engineering safety features and means for SAMG strategies in case of loss of all off-site power supply sources for VVER-1000/B320 reactor installations.•To represent important operator actions and to analyse the effectiveness of these actions for accidents management in compliance with the new approach.•The RELAP5/MOD3.3 computer code has been used in performing the analyses in a VVER-1000 Nuclear Power Plant (NPP) model. The input data deck for the analysis is optimized, verified and validated.

Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code

Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.

Loss of ‘Core cooling’ at low power and cold condition of VVER-1000/V320

This paper presents the results of thermal–hydraulic calculation of accident scenarios that involve the loss of critical safety function (CSF) “Core cooling” for VVER-1000/V320 units at Kozloduy nuclear power plant done during the development of symptom based emergency operating procedures (SB EOPs) for this plant at low power and cold condition. The main purpose of this analysis is to provide the response of monitored plant parameters to identify symptoms available to the operators and define timing for reaching the following stages during the development of processes in the reactor system: ○ Reaching the saturated temperature at the outlet of the assembly; ○ Beginning of reactor core uncovery; ○ Heating up of fuel; ○ Defining the transition time between EOPs and SAMG at temperature of 923 K; ○ Restoring of water level in the reactor; ○ Defining the CSF “Core cooling” status and the time of its loss. The results of the thermal–hydraulic analyses have been used to assist KzNPP specialists in analytical validation of EOPs at low power and cold condition. The principal acceptance criteria for EOPs are averting the onset of core damage. The RELAP5/MOD3.2 computer code has been used in performing the analyses in a VVER-1000 nuclear power plant (NPP) model. A model of VVER-1000 based on Unit 6 of KzNPP has been developed for the systems thermal–hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE – BAS), Sofia. The low power and cold condition and the modifications after the modernization program are taken into account.