Deterministic Safety Analysis Research Articles

The application study of conservative best-estimated approach for steam line break accident

The most important characteristic of main Steam Line Break (SLB) accident is the asymmetry cooldown between loops due to secondary break spurting, which results non-uniform reactivity feedback in different reactor core sections. The fuel rod DNBR limit would be more challenged in the core section with peak power. The SLB safety analysis methodologies in Administration License Application for typical PWR in China are investigated in this paper. Based on the conservative universal methodology, the best-estimated code RELAP5/MOD3 is used to perform the SLB thermal-hydraulic calculation for a three-loop PWR. In this analysis, the critical conservative models submitted for Licensing process are implemented in RELAP5/MOD3 specified modeling, which includes the Three-Channel Core Model, core inlet mixture array, core outlet mixture, vessel upper head flow and the more realistic SG model. The “second pressurizer” phenomenon in vessel upper head flashing is studied, which affects the system pressure response and mitigation. The effect of SG spouting during the transient is also studied in this paper, which is different with lumped parameter SG model. The combined approach for Deterministic safety analysis is validated in this paper, which is composed with best-estimated code, Conservative assumptions and Conservative initial & boundary conditions. Because of the flexibility and adequate conservatism, it can be spread for design optimization and independent safety assessment from the third party.

Korean perspective for integrated deterministic-probabilistic safety assessment and its synergetic strategy with conventional methods

The PSA (Probabilistic Safety Assessment) is presented as a method of comprehensively evaluating the types of accidents that can occur at nuclear power plants. One of the key components to obtain technical success was the introduction of event tree analysis, and its strength and weakness has been reviewed and enhanced with the term, for instance, Integrated Deterministic-Probabilistic Safety Assessment (IDPSA) since 1980s. However, the technical and policy maturity of IDPSA appears to have room for improvement: the terminologies need to be arranged internationally, and there are no official standards or guidelines for the method itself. Due to a misperception of IDPSA, there are even concerns that appropriate contribution in risk assessment cannot be made. For this reason, surveys and focus group interviews, sharing development experiences, and the direction of regulation and R&D on IDPSA were conducted targeting PSA experts in Korea. In order to support such process, authors have structured an overview of the development history and technical features of IDPSA. Finally, we will explore the ways to achieve synergy between the deterministic safety analysis and PSA, which may the origin of motivation how to deal with dynamic variability more properly such that an undue risk can be minimized.

Preliminary accident analysis of the loss of vacuum in vacuum vessel for the European DEMO HCPB blanket concept

Design basis accidents are investigated continuously for the European DEMO reactor accompanying its development. One selected postulated initial event (PIE) is a loss of vacuum (LOVA) in vacuum vessel (VV) with large ingress of air induced by rupture in a VV penetration. It has been investigated for the helium cooled pebble bed (HCPB) blanket concept according to the DEMO baseline 2017. The associated primary heat transfer system (PHTS) and the related systems in the tokamak building, from the VV to the PHTS vault and galleries, are considered for the investigation. The LOVA is postulated to occur at a port seal of the electron cyclotron equatorial port plug on the side of the closure plate with (i) a small leak of 1.0 × 10−3 m2, or (ii) a large break size of 1.0 × 10−2 m2. Air ingress from one port cell into the VV leads to the VV pressurization and the fusion power termination followed by an unmitigated plasma disruption. A loss of off-site power for 32 h is assumed to coincide with the disruption. An in-vessel loss of coolant accident (LOCA) is considered as a consequence if the affected first wall (FW) reaches the defined temperature of 1000 °C. The radioactive inventories in the VV (tritium, W-dust) can mobilize towards the VVPSS, the affected systems in the building and the environment due to pressurization, venting and leak conditions. MELCOR 1.8.6 for fusion is applied for this deterministic safety analysis. The resulting releases of radioactivity to the environment are then provided for dose calculation using the computer systems UFOTRI and COSYMA. Outcomes of this LOVA analysis are critically discussed: the transient evolutions of different cases are compared; hydrogen production is detected in case of aggravating FW failure; the source terms (tritium, W-dust) are transported to the connected systems; and the dose results from the environmental releases are provided.

Safety margin quantification by integrating probabilistic and deterministic safety assessments — Application to design extension conditions

Various safety analysis methods have been used over the years to quantify safety margins: at the start by Deterministic Safety Analysis (DSA) with conservative models or best estimate models and “bounding” assumptions; then by the Best Estimate Plus Uncertainty (BEPU) approach which takes into account uncertainties related to physical and modelling parameters; and recently by the Extended BEPU (EBEPU) approach with the intention of incorporating information from Probabilistic Safety Assessment (PSA).By combining probabilistic and deterministic safety assessments while taking into consideration a wide variety of uncertainties, the current work aims at the development of an Integrated Safety Margin Quantification (ISMQ) technique to extend the scope of existing approaches. Multiple aspects of the “safety margin” can be handled by this ISMQ methodology, including pertinent Initiating Events (IE) and sequences, the distance between best-estimate load and safety limit, and the likelihood of exceeding the safety limit.Together with the suggested methodology, an industrial application on Design Extension Conditions (DEC) of a typical Generation II 1000 MWe Pressurized Water Reactor (PWR) is presented to illustrate the approach and the anticipated outcomes. This ISMQ methodology is also applicable for safety analyses of other reactor designs including Small Modular Reactors (SMRs).

Preliminary thermal-hydraulic deterministic safety analysis of an in-vessel LOCA for the DTT facility

The DTT (Divertor Tokamak Test) facility is a new experimental tokamak under development by an Italian consortium in cooperation with several high-standard European laboratories. The ENEA division FSN-SICNUC is involved in the project for carrying out deterministic safety analysis of postulated accidents. In the first year of activity, it has been analyzed an In-Vessel LOCA (IVLOCA) scenario. The IVLOCA scenario selected is caused by a break in the divertor’s cassette cooling tubes with the consequent release of coolant inside the Vacuum Vessel (VV). The best estimate thermal-hydraulic system code TRACE, developed by USNRC, was selected to conduct the preliminary thermal-hydraulic analysis of this accident. DTT is currently under design, so not all the data are frozen. Therefore, based on some engineering assumptions and scaling considerations from similar facilities, the nodalization of DTT VV was developed using the three-dimensional TRACE component “VESSEL”. Then, a sensitivity analysis was carried out to simulate the break and the consequent water injection in the VV. This was done to compare the system behavior and test different nodalization approaches. Subsequently, the data of the DTT divertor cooling system were used to run additional simulations. The results allow to compare different nodalization approaches and to have a preliminary estimate of the pressure peak and temperature behavior in the VV for an IVLOCA. Finally, a first uncertainty analysis was carried out using the DAKOTA toolkit, coupled with TRACE code in SNAP. Two uncertain input parameters were selected: the rupture area of a cooling divertor tube and the temperature of the divertor coolant. The uncertainty analysis allows having a wider spectrum of system behavior and a preliminary insight on the dispersion of the VV pressure, selected as figure of merit. This paper aims to show the results of this preliminary analysis, characterizing the phenomena that occurred during the selected transient.

BEPU analysis of an Upgraded ICE facility test by TRACE/DAKOTA coupling

The Ingress of Coolant Event (ICE) in the plasma chamber is one of the safety issues in fusion nuclear plants. The best estimate thermal-hydraulic system codes adopted to perform deterministic safety analysis should be validated against the phenomena typical of accidental transients in fusion installations. TRACE (TRAC/RELAP Advanced Computational Engine), best estimate thermal-hydraulic system code developed by USNRC, has been adopted to simulate an ICE. The calculated results have been compared to the experimental data obtained in one test performed in the upgraded Integrated ICE facility at JAERI. In this updated configuration the pressure suppression system is connected to the top of the plasma chamber instead of the bottom of the vacuum vessel. The facility nodalization has been developed in the SNAP environment/architecture. To qualify the code and the nodalization, an accuracy evaluation has been performed both from a qualitative and quantitative point of view. Then, considering the presence of some uncertainties in the input-deck development, an uncertainty analysis has been carried out. The probabilistic method to propagate the input uncertainties has been selected and the analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in SNAP. In the uncertainty analysis, some relevant statistical parameters have been considered to characterize the dispersion of the results and the correlation between the uncertain input parameters selected and the PC pressure chosen as figure of merit.

Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR): An integrated nuclear-based energy production and conversion system

Based on the analysis of energy structure and the development trends of advanced nuclear energy technology, combining the complementary advantages of nuclear energy and other energy sources in China, the Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR) system has been proposed. FuSTAR system is designed to address production capacity issues in areas with energy shortages or resource concentration and can be combined with high-temperature process heat to achieve high-value-added processes such as hydrogen production. In this paper, the Tri-structural isotropic fuel combined with helical cruciform elements has been studied in detail on the neutron physics and thermal-hydraulic properties, and the flow and heat transfer experiment is carried out for the first time. The multilayer nonlinear programming method has been applied to the design of thermal transport systems. Then the optimal parameters and configuration of the thermoelectric conversion system are obtained by the superstructure optimization method. The inherent safety of FuSTAR is demonstrated through deterministic safety analysis. In addition, the iodine-sulfur cycle technology and the characteristics of tritium production are also analyzed. This study highlights the significant potential of integrating nuclear energy with other energy sources in an optimized energy-production system.

Evaluation of uncertainties associated with best estimates of parameters in the deterministic analysis of a severe accident

This paper describes an approach to evaluate the uncertainties of the results of deterministic severe accident analysis. The approach was developed in order to follow best practices of best estimate methodology and to fulfill the Russian regulatory requirements NP-001–15 in the area of atomic energy use. In accordance with these requirements, deterministic safety analyses must be accompanied with the evaluation of errors and uncertainties of the obtained results. The application of the proposed approach is demonstrated through the example of a deterministic analysis of a severe LOCA initiated by a large break at a generic nuclear power plant with VVER-1000. Within the given example a best estimate value and associated uncertainty of hydrogen mass generated at in-vessel stage is analyzed.

Validation and uncertainty analysis of ASTEC in early degradation phase against QUENCH-06 experiment

Severe Accident (SA) integral codes, such as the Accident Source Term Evaluation Code (ASTEC) developed by IRSN, are used to simulate the phenomena occurring during accident progression in Nuclear Power Plants (NPPs) up to the source term evaluation. Code validation against experimental data is fundamental to carry out deterministic safety analysis and apply these codes to NPPs. In addition, in the Best Estimate Plus Uncertainty (BEPU) framework, the quantification of the results uncertainty is needed. In the framework of the IAEA CRP I31033 “Advancing the State-of-Practice in Uncertainty and Sensitivity Methodologies for Severe Accident Analysis in Water-Cooled Reactors”, the QUENCH test-6 experiment, conducted at KIT, has been selected to develop an uncertainty analysis using the ASTEC v2.2b code. The accuracy of the best-estimate ASTEC simulation was evaluated with the Fast Fourier Transform Based Method (FFTBM) against the experimental data. Then, the uncertainty of the code results was quantified by using the probabilistic propagation of input uncertainties method, through the coupling of ASTEC with RAVEN (Risk Analysis and Virtual Environment). Beyond identifying the main sources of uncertainty affecting the simulated test, the outcomes of the work also include some general discussion on the uncertainty propagation in a SA sequence.

Simulation and design of passive residual heat removal system for Fluoride-Salt-cooled high-Temperature Advanced reactor (FuSTAR)

The new generation of reactor technology: Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR), mainly applied to the steady supply of stable electricity and high-temperature process heat for remote, underground, and arid areas. So far, the design of neutron physics and thermal hydraulics of the core have been finished, but the heat transport system and Passive Residual Heat Removal System of FuSTAR still need to be designed and optimized. Given the complexity and consistency of the design, a unified design and optimization method is required, which can directly obtain detailed configurations and dimensions. In this paper, a design and optimization method - Multi-Layer Nonlinear Programming was proposed to obtain the optimal parameters, including vital thermodynamic values, the dimensions of heat exchangers, the configurations of pipelines, and the optimal equivalent parameters in the transient simulation. Finally, based on Multi-Layer Nonlinear Programming, the effectiveness of the Passive Residual Heat Removal System was demonstrated by the deterministic safety analysis. The detailed parameters of the Passive Residual Heat Removal System in this paper can provide references for subsequent experimental verification. Moreover, for any reactor relying on a heat transport system, researchers can directly obtain its excellent preliminary configurations by this method avoiding a lot of iteration and conflict design, which is conducive to the integrated modeling and simulation of a new generation of reactors.

REactor Safety Analysis ToolboX RESA-TX

The REactor Safety Analysis ToolboX RESA-TX is a software and data package in development that combines the automatisation of all established procedures for deterministic safety analysis (DSA), the integration of expert know-how and a large database including the most relevant information required for conducting a DSA. In the current state of the art, DSA is a complex and thus error-prone process that is highly time-consuming and repetitive. The reliability of the result is strongly dependent on the availability of plant data and expert know-how. The idea of developing the RESA-TX toolbox arose at GRS to cope with these conditions. The innovative approach proposes an automated and standardised procedure, supported by a large database of plant design characteristics, plant behaviour, regulatory rules and DSA expert knowledge incorporated within the tool. Its application allows the end user to automatically generate and verify an input deck, as well as conduct design basis accident (DBA) calculations for a certain design with highly reduced manual intervention. The databases can be extended depending on available information or other boundary conditions. A heuristic approach is integrated into the model generation process, where users often suffer from a lack of information about the facility under consideration. These heuristics can be replaced when higher information quality is available or enhanced over time, which can lead to more reliable results with increasing usage of the tool. As a result, the application of RESA-TX could highly increase the efficiency of the DSA process, reducing both repetitiveness as well as user-induced errors. This in return will lead to an improvement in the quality of the analysis and reliability of the results. In consequence, RESA-TX will allow for a DSA to be conducted more frequently in situations where time or budget was a limitation before, thereby contributing to an increase in reactor safety.

Demonstration of the E-BEPU methodology for SL-LOCA in a Gen-III PWR reactor

The paper presents the first practical application of the alternative Extended Best Estimate Plus Uncertainty (E-BEPU) methodology. The E-BEPU is the systematic risk-informed combined deterministic and probabilistic methodology dedicated to design verification of Nuclear Power Plants. The approach covers applying the best-estimate computer code, realistic input data for initial and boundary conditions with uncertainties, and plant systems availability based on Probabilistic Safety Analysis methods. Furthermore, it extends the scope of the uncertainty analysis to include the availability of safety systems as an additional uncertain item. The approach is characterized by an enhanced Defence-in-Depth; it can detect cliff-edge effects and include risk insights in the deterministic analysis framework. It is an advanced approach to Deterministic Safety Analysis beyond the BEPU approaches and an alternative realization of the Extended-BEPU. In this study, the generic Generation-III Pressurized Water Reactor design was investigated to demonstrate the methodology with a practical example. The Postulated Initiating Event was a surge line double-ended break (SL-LOCA), a variant of the Large Break Loss of Coolant Accident (LB-LOCA), and considered an example of the Design Basis Event. The considered reactor design passed successfully through the procedure with a large margin due to the robustness of its safety systems.

RELAP5/MOD3.3 Analyses of Core Heatup Prevention Strategy during Extended Station Blackout in PWR

The accident at the Fukushima Dai-ichi nuclear power plant demonstrated the vulnerability of the plants on the loss of electrical power for several days, so called extended station blackout (SBO). A set of measures have been proposed and implemented in response of the accident at the Fukushima Dai-ichi nuclear power plant. The purpose of the study was to investigate the application of the deterministic safety analysis for core heatup prevention strategy of the extended SBO in pressurized water reactor, lasting 72 h. The prevention strategy selected was water injection into steam generators using turbine driven auxiliary feedwater pump (TD-AFW) or portable water injection pump. Method for assessment of the necessary pump injection flowrate is developed and presented. The necessary injection flowrate to the steam generators is determined from the calculated cumulative water mass injected by the turbine driven auxiliary feedwater pump in the analysed scenarios, when desired normal level is maintained automatically. The developed method allows assessment of the necessary injection flowrates of pump, TD-AFW or portable, for different plant configurations and number of flowrate changes. The RELAP5/MOD3.3 Patch04 computer code and input model of a two-loop pressurized water reactor is used for analyses, assuming different injection start times, flowrates and reactor coolant system losses. Three different reactor coolant system (RCS) coolant loss pathways, with corresponding leakage rate, can be expected in the pressurized water reactor (PWR) during the extended SBO: normal system leakage, reactor coolant pump seal leakage, and RCS coolant loss through letdown relief valve unless automatically isolated or until isolation is procedurally directed. Depressurization of RCS was also considered. In total, six types of RCS coolant loss scenarios were considered. Two cases were defined regarding the operation of the emergency diesel generators. Different delays of the pump injection start following the station blackout were assumed and analysed. For each scenario, two kinds of SBO calculations for two-loop PWR were performed, base and verification. Base calculations were needed to determine necessary minimum flowrate for steam generators feeding in such a way that they are not overfilled or emptied. Namely, it was assumed that instrumentation is not available during extended SBO. The verification calculations have been then performed to verify if the determined minimum flowrates are sufficient to prevent the core heatup. The calculated results show effectiveness of the proposed extended SBO prevention strategy provided that the water injection is available in the first two hours after SBO occurring at full power. If diesel generator is running after loss of offsite power for some time, e.g. one hour, the available time for steam generator water injection is significantly longer. The obtained results demonstrate the need for assessment of the pump injection flowrates before the utilization of the pump for mitigation of the event. The applicability of the developed method for assessment of the required pump injection flowrate has been validated on PWR.

Cold Leg LBLOCA uncertainty analysis using TRACE/DAKOTA coupling

Safety analyses for nuclear power plants were carried out in the past using a conservative approach. With the increase of the phenomenological knowledge, through experimental data, and computational power, it became possible to adopt best estimate thermal- hydraulic system codes to perform deterministic safety analyses. However, some uncertainties are still present in the models, correlations, initial and boundary conditions, etc. Therefore, it is fundamental to quantify the uncertainty of calculation. This approach is called “Best Estimate Plus Uncertainty” (BEPU). Among the available uncertainty analysis methodologies, the probabilistic method to propagate input uncertainty is widely adopted. In the present study, an uncertainty analysis of a cold leg large break loss of coolant accident in a generic PWR-900 MWe has been developed and it has been carried out coupling the best estimate thermal- hydraulic system code TRACE and the uncertainty quantification tool DAKOTA in the SNAP environment/architecture.

Current Status and On-Going Development of VTT’s Kraken Core Physics Computational Framework

The Kraken computational framework is a new modular calculation system designed for coupled core physics calculations. The development started at VTT Technical Research Centre of Finland in 2017, with the aim to replace VTT’s outdated legacy codes used for the deterministic safety analyses of Finnish power reactors. In addition to conventional large PWRs and BWRs, Kraken is intended to be used for the modeling of SMRs and emerging non-LWR technologies. The main computational modules include the Serpent Monte Carlo neutron and photon transport code, the Ants nodal neutronics solver, the FINIX fuel behavior module and the Kharon thermal hydraulics code, all developed at VTT. The core physics solution can be further coupled to system-scale simulations. In addition to development, significant effort has been devoted to verification and validation of the implemented methodologies. The reduced-order Ants code has been successfully used for steady-state, transient and burnup simulations of PWRs with rectangular and hexagonal core geometry. The Ants–Kharon–FINIX code sequence is actively used for the core design tasks in VTT’s district heating reactor project. This paper is a general overview on the background, functional description, current status and future plans for the Kraken framework. Due to the short history of development, Kraken has not yet been comprehensively validated or applied to full-scale core physics calculations. A review of previous studies is instead provided to exemplify the practical use.

BEPU Analysis in LBLOCA Safety Review Calculation

Deterministic safety analysis (DSA) is essential for nuclear power plant licensing. The conservative method followed CFR50 Appendix K, which will lead to a large margin. As one of the DSA methodologies, best estimate plus uncertainty (BEPU) generates more realistic results that can be used in the license application of nuclear power plants (NPPs). However, uncertainty evaluation of parameters is needed in BEPU. In this article, the safety regulatory focuses on the large break loss of coolant accident (LBLOCA) of an advanced PWR. The BEPU analysis is mainly performed by TRACE V5.0 patch 4 code, and the uncertainty analysis is conducted based on DAKOTA code. For correlation coefficients analysis, the sample size is enlarged reasonably. According to the results, this NPP meets the acceptance criteria effectively in LBLOCA with enough margin. By statistic assessment, the set of PCTs calculated has typical normal distribution characters. Based on BEPU, the uncertainties of parameters are studied. Additionally, the influence of sample size on the correlation of parameters is considered too. It could be seen that more samples could permit a more accurate estimation for Spearman partial correlation coefficient (abbreviated as SPCC). The conclusions of this article can provide technical support for the subsequent review of the safety analysis report and the design changes of NPPs.

Study on Selection of Internal Initiating Events of Small Natural Circulation Lead-Cooled Fast Reactor

The initiating events, as the starting point and basis for the deterministic safety analysis and probabilistic safety assessment of the lead-cooled fast reactor (LFR), provide important guidance on the design optimization and safe operation of reactors. This paper, based on the current design scheme of the small natural circulation lead-cooled fast reactor SNCLFR-100 and by reference to the experience of selecting initiating events of other advanced fast reactors, with the generalized “core melting” taken as the top target event, provides the internal initiating events of SNCLFR-100 by deduction in the master logic diagram (MLD) method, which form a relatively complete list of internal initiating events. This study thus can provide a theoretical basis for the safety analysis of natural circulation LFRs.

Overview of the seismic probabilistic safety assessment applied to a nuclear installation located in a low seismicity zone

Deterministic and probabilistic nuclear safety analysis methodologies have been developed and updated based on operational experience, investigation of past incidents or accidents, and analysis of postulated initiating events in order to maintain the protection of workers, the public and the environment. The evaluation of accident sequences and the total radiological risk resulting from off-site releases are general objectives addressed by these methodologies. There are hazards that continually challenge the safety of a nuclear facility or its nearby area. In particular, seismic events represent a major contributor to the risk of a nuclear facility. Different levels of ground motion induced by earthquakes may be experienced by the structures, systems and components (SSCs) of the installation. In this context, a seismic hazard analysis, seismic demand analysis and seismic fragility analysis must be carried out in order to characterize the local seismic hazard and what are the seismic demands on SSCs, allowing an adequate seismic classification of SSCs, even in installations located in sites with low seismicity. In this article, a general description of the Seismic Probabilistic Safety Assessment (Seismic PSA) methodology is presented, with emphasis on their support studies, aiming at applying the methodology described in this article to an experimental nuclear installation containing a PWR reactor designed for naval propulsion to be installed in a low seismicity zone in Brazil.

Qualification of RELAP5-3D code condensation model against full-scale PERSEO Test 9

Passive systems are being implemented in several nuclear plants already constructed, under construction or under design. The purpose of these systems is to increase the inherent safety of the plant in any transient condition, so they can provide a higher safety level in case of an accident. Best estimate thermal-hydraulic system codes, widely adopted in the nuclear field to perform deterministic safety analyses, need to be validated against the phenomena typical of passive systems (e.g. natural circulation, thermal-hydraulic behavior of large pools, low pressure phenomena, etc.). An experimental test facility called PERSEO (in-Pool Energy Removal System for Emergency Operation) was built at SIET laboratory in Piacenza (Italy) modifying the existing PANTHERS IC-PCC facility. It is a full-scale test facility aimed at studying a new passive heat removal system and it can be used to analyze the capability of system codes to simulate the related phenomena. In the present analysis, Test 9, which was performed in the PERSEO facility at around 4.1 MPa to study the activation and the heat removal capability of the proposed system, has been simulated with the RELAP5-3D code to assess its capability to predict the dominant phenomena of the transient. Particular attention has been dedicated to the analysis of condensation heat transfer coefficient computed by the code, since in previous analyses it has been found an evident underestimation of the power exchanged by the passive system. Alternative correlations for the condensation heat transfer coefficient have been identified and a correction factor for the heat transfer coefficient calculated by RELAP5-3D has been computed. After the application of the correction factor, the set composed by Nusselt-Kutateladze-Chen correlations provided the better agreement with the experimental data. Finally, the accuracy of the code calculation has been evaluated qualitatively and quantitatively by adopting the Fast Fourier Transform Based Method.

Методология анализа внутренних воздействий на безопасность АЭС

The basis for safety operation of NPPs are laid down at the design stage, therefore the main tasks of this stage are the most complete consideration of safety requirements and principles in the project. The existing experience in the design of nuclear power plants for foreign customers shows that both safety standards and national regulatory documents of a number of countries impose detailed requirements for deterministic safety analysis for internal hazards, the source of which is inside the building: fires, explosions, missiles, heavy load drop, flooding, etc. At the same time, methodological questions of performing deterministic safety analyzes for internal hazards are not sufficiently developed either at the level of national regulatory, or at the level of safety guidelines. The presented analysis methodology is aimed at improving the regulatory documentation for the safety justification of nuclear power plants and is developed on the basis of regulatory requirements in the field of nuclear safety in various countries and summarizing the experience of performing deterministic safety justification of domestic and foreign projects.