Critical Heat Flux Ratio Research Articles

Design and analysis of KA-130, the marine-based SMR for icebreakers

Designed for icebreakers, the KA-130 integral SMR provides a carbon-free solution for navigating the North Pole Route with nuclear propulsion. As an integral reactor, all primary components are integrated into the reactor pressure vessel, eliminating large-break loss-of-coolant accidents and enhancing reactor safety. The study evaluates its load following capability and demonstrates efficient load changes within two minutes. The safety analysis during small break loss-of-coolant accidents (SBLOCA) incorporates ship movements into the minimum critical heat flux ratio (MCHFR) calculations and includes a sensitivity analysis to identify significant scenarios. The most limiting case was the guillotine break at the safety injection line with low initial pressure, high core exit temperature and bottom-peaked core. The analysis shows that core flooding is maintained, and decay heat removal is successfully performed. Acceptance criteria, aligned with regulatory and KA-130 specific standards, ensure safety during and after SBLOCA, meeting MCHFR and collapsed water level conditions above the core.

Nuclear safety operating margin using new power peaking factors for NUR nuclear research reactor

Among a series of important safety parameters necessary for a complete analysis of a nuclear reactor core, the nuclear safety operating margin of a nuclear reactor is of great importance. The methodology generally used is to combine thermalhydraulic calculations with neutronic calculations. In this work, new Power Peaking Factors (PPFs) are calculated for the NUR nuclear research reactor core X-1 configuration by two ways; a deterministic way using WIMSD-CITVAP neutronic calculation in a fine calculation model, and a second way by a probabilistic method using MCNP5 model. The obtained results give the PPFs for each nuclear reactor core channel and the hot channel is then determined. Once this channel known, an asymptotic thermalhydraulic model is applied. This model gives the coolant and the cladding temperature evolutions in the hot channel for the calculated PPFs in the two presented models. All these results are compared with those obtained using the old PPF commonly used in such calculations. Results show that PPFs obtained by deterministic method are the most conservative. Thus, this last result is retained for a nuclear safety evaluation. An equation set-up is made for calculating the critical heat flux ratio. This ratio is verified to be greater than the limit of nuclear safety operating margin for NUR nuclear reactor.

Evaluation of the success of in-vessel corium retention using external reactor vessel cooling mechanism for a small PWR

Among the several study fields of severe accidents, In-Vessel Retention (IVR) via External Reactor Vessel Cooling (ERVC) referred to as IVR-ERVC is a critical subject being explored for various reactor designs. This study aims to investigate and assess the effectiveness of IVR of molten corium in a PWR utilizing the ERVC mechanism. A RELAP5/SCDAP model developed for the assessment of IVR-ERVC was applied to a small PWR after its successful validation. The boundary conditions applied to the model are evaluated through simulation of severe accidents of the small reactor (998 MWth). Thermal hydraulic studies of sensitive parameters have been carried out to determine the safety margin and effectiveness of the IVR-ERVC method under a variety of process conditions. The results indicate that a high mass flux is required to avoid boiling crises at low degrees of subcooling (around saturation conditions). A mass flux of 186.44 kg/m2-s or higher is required to provide a sufficient safety margin for inlet water with degree of subcooling of 8 K. The minimum Critical Heat Flux (CHF) is found to be 2.5 MW/m2 at a mass flux of 186.0 kg/m2-s for this case. Critical Heat Flux Ratio (CHFR) is found in the range of 6.0–17.0 with the cooling water flow rate of 20 kg/s (mass flux of 34.48 kg/m2-s) and 65 K subcooling, which provides a sufficient safety margin to avoid boiling crisis. It is concluded that the ERVC is an effective mechanism for the prevention of RPV rupture under active and passive conditions of the reference reactor.

Experimental and analytical natural circulation analysis of the Jordanian 5 MW research reactor using plate type fuel by RELAP5 MOD3.3

The 5-MWth Jordanian research reactor is an open-pool type research reactor using plate-type fuel. It is cooled by the upward natural flow at low power-operation. Natural circulation cooling is important in the safety analysis because it is the last scenario of all accidents. Therefore, studying natural circulation is important in determining the safety limits, such as the maximum allowed temperature and the minimum critical heat flux ratio of fuel. In this paper, the reactor operated at 50KW by the natural circulation to investigate its neutronics and thermal–hydraulic characteristics. Then, the results were benchmarked with RELAP5 MOD 3.3 by implementing models with and without considering the effect of heat loss through the Upper Guide Structure. As a result of this study, different models show minor differences in thermodynamic parameters and an additional safety margin. Eventually, benchmarking the results and the calculations of RELAP5 led to a reasonable agreement in both simulated models.

Steady-State Subchannel Thermalhydraulic Assessment of a Full-Scale Pressurized Water Reactor-Small Modular Reactor Fuel Assembly With Conventional and Advanced Fuels

Abstract Thermalhydraulic analyses using subchannel codes (e.g., ASSERT-PV) are performed as a support tool to evaluate safety margins and the key parameters. Advanced fuels have recently attracted the international community's interest to improve safety margins during normal operation and accident scenarios by utilizing special coatings and barriers in a heterogeneous, multiregion, multicoating, multiclad annular fuel. In addition, advanced fuels improve the performance characteristics such as a higher burnup and better uranium utilization. Therefore, there is a need to understand the implications of such advanced unconventional fuels for the landscape of the Canadian nuclear industry and Canadian policy for energy development. In this work, subchannel thermalhydraulic analysis of a small modular reactor based on pressurized water reactor technology (PWR-small modular reactor (SMR)) core is performed using ASSERT-PV. A benchmark of a 17 × 17 fuel assembly with conventional fuel elements in comparison to PWR-SMR specification was conducted. The advanced fuel element system is also investigated and compared with the conventional one. The results indicated that the advanced fuel achieves a significant reduction in fuel element temperature by 15%. In addition, the results revealed that the proposed advanced fuel could achieve a minimum critical heat flux ratio (MCHFR) higher than the conventional fuel by 17%.

Simulation of Zircaloy and SiC cladding thermal performance of a nuclear fuel rod subassembly

A thermal-hydraulic analysis of a nuclear fuel rod subassembly of a Pressurized Water Reactor (PWR) has been performed during a Loss of Flow Accident (LOFA). Two different cladding materials are studied in this exercise namely, (a) conventional Zircaloy (Zr) and (b) a Silicon Carbide (SiC) cladding concept approach using monolithic SiC and SiC composite two-layer structure. The simulation is carried out by using USNRC licensed and AERB approved RELAP5/MOD3.4 code. A short-time problem is analyzed, neglecting the burnup change and irradiation effects. A transition into film boiling is observed in the upper part of the rod during LOFA, causing the convective heat transfer coefficient to decrease. The peak cladding temperatures and the Critical Heat Flux Ratio (CHFR) values of the two cladding materials are compared. From a safety perspective, the use of SiC is considered as it will be able to maintain performance during the operational transients.

Thermal-hydraulic analysis of light water reactors under different steady-state operating conditions, Part 1: Boiling water reactor

The steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA. The analysis' goal is to keep the thermal safety margin under control and the core integrity intact under steady-state operating conditions. The effects of operating conditions such as power distribution, power level, and coolant mass flow rate on the pro- posed core's performance are investigated. For this purpose, the one-dimensional computer code MITH was used. The code's reliability was tested using the General Electric benchmark 3579 MW reactor. Two-channel models were tested (the average and the hot channel). Ther- mal-hydraulic parameters such as fuel-centerline, fuel-surface, outer clad surface and coolant temperature, critical and actual local heat flux, critical and minimum critical heat flux ratio and pressure drop are evaluated along the tested channels. Temperatures, as well as actual and critical heat flux distribution profiles, were obtained. The tested operating conditions had a significant influence on these parameters, and also on the thermal-hydraulic performance. The obtained results are in good agreement with the data from the tested core. The obtained results are well within the safety margins. The good agreement between tested reactor data and MITH code calculation concerning the reactor demonstrates the reliability of the analysis methodology from a thermal-hydraulic perspective.

Conservative analytical and experimental investigation of all pumps failure event in the primary cooling system for a 5 MWth research reactor

The 5-MWth Jordanian Nuclear Research Reactor is an open-tank and open-pool research reactor which uses narrow rectangular fuel. The reactor is cooled by demineralized water in downward forced flow and natural circulation in transient events. The reactor performed a loss of flow test in the Primary Cooling System (PCS) by turning off all PCS pumps at full-power operation. The reactor then displayed safe behavior in thermal-hydraulic, neutronics, and acceptable radiological values. Besides, the RELAP5 MOD3.3 code was used to implement the reactor in three different comprehensive models according to several safety analysis conservatisms, such as assuming a lower total core flow rate by −12% or assuming failure in a safety valve. After that, a benchmark was conducted between the experimental results, and the code calculations of implemented models showed well agreement in general thermal-hydraulic and safety analysis parameters. The code proved a safe reactor trend that used all conservative assumptions even in the most conservative implemented model. The primary safety parameters, the Maximum Temperature of Fuel Clad (MTFC) and the Minimum Critical Heat Flux Ratio (MCHFR), were 421 K and 1.82 for a short time, respectively. The resulting parameters underwent the event's acceptance criteria, where the limits of the set values are 673 K and 1.5 for MTFC and MCHFR, respectively. As a result of the study, several models illustrated differences in thermodynamic parameter effects and gave higher and additional reactor safety levels. Moreover, the test comprehensively investigated the reactor behaviors and ensured the reactor cooling system capability by free convection cooling during the event.

Thermal-hydraulic modelling and analysis of ITER tungsten divertor monoblock

• A mathematical model has been developed to investigate the thermal-hydraulic performance of ITER tungsten divertor monoblock. • The model predicts the thermal response of the divertor structural materials and coolant tube. • The model accounts for the melting, vaporization, and solidification of the upper layer of the divertor facing plasma. • The model is verified against a previous calculation in the literature for DEMO divertor at an incident surface heat flux of 10 MW/m 2 . • The model is used to simulate the divertor materials response subjected to a VDE. The divertor is a fundamental component of fusion power plants, being primarily responsible for power exhaust and impurity removal via guided plasma exhaust. Due to its position and functions, the divertor has to sustain very high heat flux arising from the plasma (up to 20 MW/m 2 ), while experiencing an intense nuclear deposited power, which could jeopardize its structure and limit its lifetime. Therefore, attention has to be paid to the thermal–hydraulic design of its cooling system. In this work a mathematical model has been developed to investigate the steady state and transient thermal–hydraulic performance of ITER tungsten divertor monoblock. The model predicts the thermal response of the divertor structural materials and coolant tube. The coolant tube is divided into specified axial regions and the divertor plate is divided into specified radial zones, and then a two-dimensional heat conduction calculation is performed to predict the temperature distribution for both steady and transient states. A two-dimensional numerical finite difference technique is adapted in Cartesian coordinate system where the implicit scheme is used for transient calculation. The model also accounts for the melting, vaporization, and re-solidification of the upper layer of the divertor facing plasma. The selected heat transfer correlations cover all possible operating conditions of ITER under both normal and off-normal situations. The model is verified against a previous calculation in the literature for DEMO divertor at an incident surface heat flux of 10 MW/m 2 . The model is then used to predict the steady state thermal behaviour of the divertor under incident surface heat fluxes ranges from 2 to 20 MW/m 2 for a bare cooling tube and cooling tube with swirl-tape insertion. It calculates the maximum tube surface heat flux and the minimum critical heat flux ratio for all cases as well. The model is also used to simulate the divertor materials response subjected to high heat flux during a vertical displacement event (VDE) where 60 MJ/m 2 plasma energy is deposited over 500 ms.

Optimization of AHWR core design: neutronics and thermal-hydraulics analysis and calculations

Advanced Heavy Water Reactor(AHWR) is being designed by BARC for thorium utilisation and demonstrate all aspects of the thorium fuel cycle. The AHWR is a vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The equilibrium fuel cycle of AHWR is being optimised for a discharge burn-up of 40 GWd/Te with Plutonium as makeup fuel in a closed fuel cycle. The 233U content in the Uranium and fissile Plutonium content in the Plutonium have been assumed to be about 78% and 75% respectively. One of the important passive safety features of AHWR is heat removal through natural circulation which is governed by a strong neutronic and thermal hydraulic coupling. The neutronics design simulations were done for average coolant condition. In order to study the effect of natural circulation and its consequences on the coolant behaviour and the resulting power distribution, the neutronics and thermal–hydraulic coupled calculations were carried out. The neutronics calculations code (FEMFOL) and thermal–hydraulic calculations code (ARTHA) were externally coupled to estimate the coolant density distribution and the critical heat flux ratio (CHFR) for the fuel. This paper gives the details of the critical heat flux ratio estimation and the effect of density distribution on the core power distribution at some selected operating phases during core follow-up of AHWR.

Automated critical power limit estimation for natural convection-cooled research reactor core

The critical power limit for a natural circulation-cooled research reactor core is estimated by utilizing a system analysis code RELAP5/MOD3.3 and an in-house companion program RCPP. The variation range of thermal-hydraulic parameters is found from steady-state and transient simulation results. From literature, potential candidate correlations for critical power estimation were selected and the design limit critical heat flux ratio was evaluated for the chosen correlations based on error statistics. A computational fluid dynamic simulation was carried out to observe the power-equilibrium flow relationship and used to prepare the system code input. To ease a rigorous critical power estimation process, an in-house code was developed to automate batch run and output post-processing of system code. The critical power limit was found for each combination of major operation parameters. A critical power map was generated by merging each power limit value, and the effect of the correlation and core states were studied.

Verification of the TIGRSP Computer Code as Applied to a 19-Rod Fuel Assembly Using CFD Computations

For assessing the core critical heat flux ratio in VVER-type nuclear reactors, the lattice computation code TIGRSP (steady-state thermal and hydraulic analysis of parameters in fuel rod bundles) is used. Despite the fact that quite a large amount of verification studies has been accomplished and that the code has passed the certification procedure, work on improving the code capabilities is still being continued. One of the modern tendencies is the wide-scale involvement of CFD codes for studying local parameters in fuel assemblies (FAs). Along with comparison with experimental data, it is advisable to carry out cross verification of lattice codes with the results of 3D thermal-hydraulic computations for conditions for which experimental data are either lacking or limited for some or other reasons. This article considers thermal-hydraulic processes during coolant flow in a 19-rod assembly with parameters close to the nominal conditions in the FAs of a VVER-1000 reactor. The 3D model is verified with respect to empirical formulas and experimental data. Based on the results of test and verification computations, we determined the mesh and turbulence model parameters using the ANSYS CFX code by means of which it is possible to obtain reliable results in estimating the hydraulic characteristics of FAs fitted with spacer grids, heat transfer in the single-phase region, and coolant flow distribution over the FA cross section. The CFD computation of coolant flow for the 19-rod assembly is carried out, and the coolant mass velocity and temperature in different sections over the length and corresponding cells of the TIGRSP code are calculated. The TIGRSP code has been verified against experimental data and taking into account the CFD simulation of the 19-rod FA, and a software module for graphically visualizing the lattice code output results has been developed. The presented data can be used for verification of thermal-hydraulic codes and in elaborating the design of FAs for VVER-type reactors.

Safety margin characterization for high power fueled experiments

The updated safety basis of the Advanced Test Reactor (ATR) has specific criteria for experiments that are cooled by the primary coolant. It requires that during a Condition 2 transient (where flow coastdown is typically the most limiting), the Critical Heat Flux ratio (CHFR) is greater than 2.0 after consideration of several power multipliers. This safety criteria is often overly conservative for fueled experiments and restricts the acceptance of experiments. In the updated safety basis, a provision for a stochastic approach is included. In this study, coupled Dakota-RELAP5 and RAVEN-RELAP5 are used to propagate uncertainties. The objective of this study is to provide a quantitative basis for utilizing the provisions in the ATR experiment safety criteria.Three different fueled experiments, MP-1, FSP-1, and AFIP-6 are considered. RELAP5 models for each experiment are used to simulate the flow coastdown transient. Several thermal-hydraulic correlations are assessed to determine the Onset of Significant Voiding (OSV), Onset of Flow Instability (OFI), and Critical Heat Flux (CHF). The OSV and OFI correlations considered indicate that there is a large safety margin. The CHF correlations also indicate that there is a moderate to large safety margin. The high mass flux of the experiments exceed the Look-up Table (LUT) correlation range. This results in an artificially restricted prediction of CHF value. A high mass flux modifier was implemented to improve LUT applicability. Due to the specifications of the experiments considered, the correlations must be applicable to higher mass flux ranges. It was also found that, often, the conventional CHFR >2.0 criterion is not met. However, the statistical CHFR margin (μ−3σ>1.2) is met. This result suggests that new criteria, while maintaining adequate safety margins, can be used to expand the operating envelope of high power fueled experiments at ATR. The Pearson ranking indicates that the power level and mass flux have the largest impact on the CHFR and OFI predictions. If a reduction of uncertainty in predictions is sought, the factors affecting fuel experiment fission power should be prioritized for scrutiny, followed by the mass flow rate.

Analysis of a loss-of-flow accident resulting from the primary pump shaft break transient of the NUR research reactor

ABSTRACT This paper presents a combined experimental and numerical simulation approach to investigating safety issues related to postulated loss-of-flow accident (LOFA) cases, which are more likely to occur in the NUR Research Reactor (Algeria). The transients investigated at nominal-power operating conditions are related to the loss of flow resulting from an instantaneous shaft break in the main cooling pump of the NUR reactor. The investigations are based on hydrodynamic and thermal hydraulic experiments to assess the reactor cooling system’s behavior. 3D Monte Carlo neutron transport calculations were performed with the (MCNP) code to determine the resulting neutronic properties of the core. In the accident analysis, a model of the primary cooling system was applied via the RELAP5 code. The experimental data and RELAP5 predictions showed good agreement. Additionally, the LOFA due to the transient scenario of the pump shaft break was compared with the LOFA due to normal loss of the coolant pump power transient. The results obtained from the transient (LOFA) studies revealed that in both cases, the lower limit of the minimum critical heat flux ratio and minimum onset of flow instability ratio for NUR is satisfied with a sufficient margin.

Steady-State Subchannel Thermal-Hydraulic Assessment of Proposed Thorium Fuel Designs

The Canadian Nuclear Laboratories (CNL) is developing the technologies to enable the use of thorium-based fuels in pressure tube–heavy water reactors (PT-HWRs). One of the key stages in developing the thorium-based fuels for PT-HWRs is the reactor core configuration. Currently at CNL there are 20 core configurations under investigation, which involve several types of thorium-based fuels that could be implemented in a 700-MW(electric)-class PT-HWR. Among these core configurations, four fuel bundle concepts are being considered: (1) the reference (or nominal) 37-element bundle; (2) a 37-element modified bundle, with the center element using a different fuel material; (3) a 35-element bundle; and (4) an 18-element internally cooled annular fuel bundle. This study presents the steady-state subchannel thermal-hydraulic assessment of the 20 core configurations under investigation. The hottest channel approach is used in this study, as it represents the upper limit of a feasible design. The axial and element power distributions used in the analysis correspond to those of the discharge burnup. Three mass flows are considered in this study: 13.5, 21, and 24 kg/s. Five parameters are used to evaluate the fuel channel/bundle performance, namely, minimum critical heat flux ratio, channel pressure drop, enthalpy distribution, void fraction, and core power.

Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

Abstract In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1X-A1, U3O8-Al, and U3Si2-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1X-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U3Si2-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

Neutronics and Safety Studies on a Research Reactor Concept for an Advanced Neutron Source

This paper presents preliminary neutronics and thermal hydraulics safety analysis results for a low-enriched uranium (LEU) fueled research reactor concept being studied at the National Institute of Standards and Technology (NIST). The main goal of this research reactor is to provide advanced sources for neutron scattering experiments with a particular emphasis given to high intensity cold neutron sources (CNSs). A tank-in-pool type reactor with an innovative horizontally split compact core was developed in order to maximize the yield of the thermal flux trap in the reflector area. The reactor concept considered a 20 MW thermal power and a 30-day operating cycle. For non-proliferation purposes, a LEU fuel (U3Si2-Al) with 19.75 wt% enrichment was used. The core performance characteristics of an equilibrium cycle with several representative burnup states—including startup and end of cycle—were obtained using the Monte Carlo–based code MCNP6. The estimated maximum perturbed thermal flux of the core is ~5.0 × 1014 n/cm2-s. The calculated brightness of the CNS demonstrates an average gain factor of ~4 compared to the current source operated at the existing NIST reactor. Sufficient reactivity control worth and shutdown margins were provided by hafnium control elements. Reactivity coefficients were evaluated to ensure negative feedback. Thermal hydraulics safety studies of the reactor were performed using the multi-channel safety analysis code PARET. Steady-state analysis shows that the peak cladding temperature and minimum critical heat flux ratio are less than design limits with sufficient safety margins. Detailed transient analyses for a couple of hypothetical design-basis accidents show that no fuel damage or cladding failure would occur with the protection of reactor scrams. All these study results suggest this new research reactor concept offers a demonstrable potential to greatly expand the cold neutron capability with a 20 MW power and certified LEU fuels.

Uncertainty assessment for loss of flow accident of a 5 MW pool-type research reactor

Best estimate plus uncertainty (BEPU) is a promising approach to the safety analysis of nuclear reactors, and the uncertainty calculation is a very important concern about it. BEPU ensures realistic safety margins and secures higher reactor effectiveness by adopting best-estimate codes and realistic input data with uncertainties, whereas the previous conservative analysis generates excessive conservatism by considering each input parameter separately. A loss of flow accident (LOFA) of a 5MW open-pool type research reactor was selected as a sample problem for a BEPU uncertainty assessment. We selected the failures of all primary cooling system (PCS) pumps, which would cause the abrupt reduction of flow and the reversal of core flow. The significant contributors to the reactor safety were identified and then input sets were sampled. For the uncertainty evaluation, 124 calculations were performed. This is the number of code runs required for a 95%/95% level with the 3rd order Wilk’s formula. The MOSAIQUE software developed by Korean Atomic Energy Research Institute (KAERI) was used for automated sampling of the uncertainty parameters, a global uncertainty calculation, and post processing of the results. The critical heat flux ratio (CHFR) and the fuel centerline temperature (FCT) were calculated at the 95%/95% level and were compared with those from conservative analyses. In addition, the impact of each design variables on the safety parameters was estimated by sensitivity analysis.

Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

Abstract Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 × 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties ...

Study of reactor plant disturbed cooling condition modes caused by the VVER reactor secondary circuit

Based on the equations of energy, mass and momentum conservation for the three-dimensional unsteady flow of the two-phase mixture, simulations were made of reactor plant disturbed cooling condition modes caused by the VVER secondary circuit using computer thermohydraulic codes. The use of computer codes to investigate these modes makes it possible to analyze transient and some emergency processes, without involving the industrial testing method; this helps create a basis for solving the problems of reliability, operational safety, and efficiency of nuclear power plants.A modern nuclear reactor is a complex system which cannot be investigated or calculated by using only simple theoretical models. Thermohydraulic calculations are an essential part of most of the design and technological developments in the nuclear power industry. Since the NPP conditions exclude the traditional technical way to check and specify the results and conclusions of an aprioristic analysis based on industrial testing, in some instances, it is possible to examine and predict the parameters of thermohydraulic processes in the reactor circulation circuit involving computer simulations.The primary objective of this work is to calculate and investigate reactor disturbed plant cooling condition modes caused by the VVER secondary circuit in order to determine whether the calculated parameters conform to the acceptance criteria established in the regulatory documents.Based on the RELAP-5, TRAC, and TRACE software codes, reactor plant cooling condition malfunction modes caused by the VVER-1000 secondary circuit were simulated and investigated. Experimental data on the mode with the turbine-generator stop valve closing are presented. The obtained dependences made it possible to determine the maximum values of pressure and temperature in the circulation circuit as well as estimate the Minimum Critical Heat Flux Ratio (MCHFR). It has been found that, if any of the initial events occurs, safety systems are activated according to the set points; transient processes are stabilized in time; and the Critical Heat Flux (CHF) limit is provided. Therefore, in the event of emergency associated with the considered modes, the reactor plant safety will be ensured.