Safety Analysis Report Research Articles

Conceptual Design, Development, and Preliminary Safety Evaluation of a PWR Dry Storage Module for Spent Nuclear Fuel

Dry storage systems are one of the storage methods for spent nuclear fuel used in many countries that operate nuclear power plants. To ensure the safe storage of spent nuclear fuel, dry storage systems are designed to maintain radiation shielding, thermal management, and subcritical and mechanical integrity. In addition, these systems must be able to withstand earthquakes, tornadoes, floods, extreme temperatures, and other operating and design-based accident conditions. In order to develop a model with safety and economic feasibility by analyzing various dry storage systems, a vertically dry storage module was proposed, and evaluations were performed on safety evaluation along with design requirements. As a result of the evaluation, all of the safety design requirements were met. This evaluation’s results can be used as basic data for the detailed design of the dry storage module to proceed with further research, including the preparation of a safety analysis report and experimental verification for licensing applications.

Preliminary Evaluation of Derived Concentration Guideline Level for Surface Soil at Wolsong NPP Site Using RESRAD-ONSITE Code

A preliminary derived concentration guideline level (DCGL) was calculated for the soil of the Wolsong nuclear power plant (NPP) (the first commercial pressurized heavy water reactor in South Korea) site using the RESRAD-ONSITE computational code. In total, fourteen selected radionuclides were analyzed after considering the preliminary evaluation information of radionuclides observed in the pressure tube specimen in the Wolsong Unit 1 heavy water reactor and previous NPP decommissioning cases. Furthermore, a geological structure model of the Wolsong NPP site was established according to the safety analysis report of the Wolsong NPP. In addition, the distribution coefficients (Kd) of various radionuclides were derived using the JAEA-SDB and the pH information of the groundwater around the Wolsong NPP site. The DCGL for surface soil of the Wolsong NPP site was derived via the application of criteria for the site release to facilitate unrestricted reuse. Moreover, preliminary dose evaluation and relevant analysis were performed according to the Wolsong NPP site resident scenario. The novelty of this study lies in the first calculation of the preliminary DCGL values for the case of the pressurized heavy water reactor (Wolsong NPP) site. It is expected that further reliable DCGL results might be achievable if more precise radionuclide information and site-specific parameters with respect to the Wolsong NPP site are secured and applied in the future.

Fractional order multi-scheduling parameters based LPV modelling and robust switching H∞ controllers design for steam dump system of nuclear power plant

In this research work, the highly challenging problem of novel modelling and nonlinear control of steam dump system of Pressurized Water Reactor (PWR) type Nuclear Power Plant (NPP) is attempted. The Fractional Order Multi- Scheduling Parameters based Multi-Input Single- Output Linear Parameter Varying (FO-MSP-MISO-LPV) model of Steam Dump System (SDS) is estimated with uncertain dynamics under sudden load variation transients. MSP for uncertain dynamics of SDS in FO framework is the most challenging problem and attempted in a novel fashion for the first time in nuclear industry. Scheduling parameters are dynamic in nature that makes the control problem more challenging. The Model is estimated experimentally by least square method using innovative plant operational data of opening positions of different valves as input variables and steam pressure as an output variable and cold leg coolant temperature coefficient of reactivity, hot leg coolant temperature coefficient, steam flow rate and turbine power as dynamic scheduling parameters. A switching controller is designed to address variable conditions of steam pressure for the actuation of dump valves, relief valves and safety valves in SDS. A robust fractional order LPV switching H∞ (RFO-LPV-SWH∞) controllers are formulated and designed for FO-MSP-MISO-LPV model. The design of RFO-LPV-SWH∞ controllers is another significant contribution in switching mode with non-integer and LPV hybrid framework. RFO-LPV-SWH∞ controllers are tested, simulated and validated against benchmark transients as laid down in Final Safety Analysis Report (FSAR) of PWR-type NPP. The input and output variables at first and second vertex of polytope are fast reference tracking under highly nonlinear uncertain dynamics of SDS. Closed loop simulation experiments are conducted and proved that the proposed closed framework is robust in performance under parametric uncertainty.

Validation of the thermal-hydraulic thunder code and application for PWR core in transient conditions

The THUNDER computer code allows performing thermal-hydraulic core analysis of pressurized water reactors aiming at the design and safety analysis of power and naval reactors in steady state and transient conditions. It uses the subchannel and control volume methods to perform a tridimensional analysis of the nuclear reactor core. The conservation equations are solved simultaneously for each control volume based on geometric and operation conditions. The Integral Momentum Method is used for solving the conservation equations decoupling the momentum equation from the mass and energy conservation equations. The drift-flux model is adopted to treat the two-phase flow condition. The analysis is performed in two-stages: first, a macroscopic analysis is conducted with subchannel equivalent to a fuel assembly (bundle), and second, with a subchannel equivalent to an individual coolant channel. Only the hottest fuel assembly is analyzed with detail. To validate the THUNDER code we used the experimental data from the NUPEC PWR benchmark (subchannel and bundle tests), in which we verify the accuracy of the coolant temperature at steady-state in 4 different void fraction conditions and the accuracy of the void fraction at transients initiated by four different conditions: core power increase, reduction of inlet mass flow rate, increase of inlet coolant temperature and depressurization of the inlet coolant. At steady-state conditions, the mean relative discrepancy for the coolant temperature was 0.09. At transient conditions, the void fraction mean absolute discrepancy for power increase was 0.045, for mass flow reduction was 0.079, for inlet coolant depressurization was 0.032 and for temperature increase was 0.092. The code was applied for the Brazilian 1350 MW nuclear power plant ANGRA 2 for evaluating the thermal-hydraulic safety parameters at steady-state, as well as the operational transients of loss of one or several primary coolant pumps. The THUNDER code results showed that the safety limits were not exceeded in concordance to the Final Safety Analysis Report of the plant. These results indicate that the THUNDER code is sufficiently accurate to perform thermal-hydraulic analyses for PWR power plants at steady-state and transient conditions.

DEMO – The main achievements of the Pre – Concept phase of the safety and environmental work package and the development of the GSSR

The EUROfusion Safety and Environmental Work Package (WPSAE) has the scope to progress the safety studies for the future EU DEMO reactor. A Generic Site Safety Report (GSSR) has been prepared to include all the steps necessary to cover the safety issues of the nuclear fusion plant, from the definition of principles and requirements, through the detection of the source terms at risk, selecting the postulated initiating events, analyzing the accidents, quantifying the doses to the population and investigating waste production and its management. Eleven GSSR volumes collate the studies performed. The final goal is to prepare safety documentation, as complete as possible, to initiate a preliminary safety report when the plant site is selected.In parallel, the safety studies are supporting the maturation of the design of DEMO, providing feedback on the technical choices for the machine, the selection of materials, the use of space, the equipment necessary to correctly manage the safety risks through continuous collaboration with the design teams of other DEMO Work Packages.The GSSR focused on safety and environmental issues, it did not encompass security issues that could impact the design and hence the safety performance of DEMO. Consideration of the interactions between safety and security will be needed in the development of the DEMO detailed design and its associate Preliminary Safety Analysis Report (PSAR).The main achievements of the Pre-Concept Phase are presented in this paper, by summarizing the contents of GSSR volumes. Completion of the work is foreseen during the future Concept Design Phase (2021–2027).

Heat balance analysis of RSG-GAS coolant system due to cooling tower refurbishment

The RSG-GAS reactor is designed to operate at a power level of 30MW, as stated in the Safety Analysis Report, and currently RSG GAS has been operating for 34 years. Aging of components has been anticipated to ensure the reactor operates safely. Some parts of the structure, system and components (SSC) will experience degradation/aging and must be replaced. Cooling tower is one of them, where the sub components are maintained and tested regularly. Cooling tower aging indication are known from the decreased performance of the reactor cooling system. For this reason, it is necessary to evaluate, both in terms of measurement and calculation analysis, before and after the refurbishment. The efficiency of the cooling tower has been calculated based on the measurement results of the cooling temperature, at the RSG-GAS operation at a power level of 30 MW. From the measurement results, it can be seen that there is a difference in air velocity inlet and outlet of module 1 and module 2. As well as the working temperature of the blower is uneven, the blower temperature ranges from 60.5°C-91.9°C. So that the operation of the RSG GAS could not last long. The efficiency of the cooling tower during the day ranges from 54.72% to 73.23%. The effect of decreasing cooling tower capability can be seen from the calculation results on the performance of each module individually. while its effect on the overall performance is known from the increase in the temperature of the coolant entering the reactor from the HE. The results of the cooling temperature measurement, after refurbishment, show that the cooling tower performance has been able to dissipate heat according to the initial design.

Analysis of the anticipated transient without scram (ATWS) initiated by emergency power mode through the full scope simulator

Abstract The integrity of the reactor coolant system is severely challenged as a result of an Emergency Power Mode – ATWS event. The purpose of this paper is to simulate the Anticipated Transient without Scram (ATWS) using the full scope simulator of Angra 2 Nuclear Power Plant with the Emergency Power Case as a precursor event. The results are discussed and will be used to examine the integrity of the reactor coolant system. In addition, the results were compared with the data presented in Final Safety Analysis Report (FSAR – Angra 2) in order to guarantee the validation of the methodology and from there analyze other precursor events of ATWS which presented only plausibility studies in FSAR – Angra 2. In this way, the aim is to provide and develop the knowledge and skill necessaries for control room operating personnel to ensure safe and reliable plant operation and stimulate information in the nuclear area through the academic training of new engineers. In the presented paper the most severe scenario is analyzed in which the Reactor Coolant System reaches its highest level of coolant pressure. This scenario is initiated by the turbine trip jointly with the loss of electric power systems (Emergency Power Mode). In addition, the failure of the reactor shutdown system occurs, i.e., control rods fail to drop into the reactor core. The reactor power is safely reduced through the inherent reactivity feedback of the moderator and fuel, together with an automatic boron injection. Several operational variables were analyzed and their profiles over time are shown in order to provide data and benchmarking references. At the end of the event, it was noted that Reactor shutdown is assured, as is the maintenance of subcriticality. Residual heat removal is ensured.

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.

An integrated method for atmospheric dispersion and corresponding risks: Application to ITU TRIGA Mark II research reactor

The prediction of the present and future impact of radioactivity release is important. However, there are several models and options to choose from. An integrated approach is proposed in this study and it is implemented to Istanbul Technical University (ITU) TRIGA Mark II Research Reactor. The procedure was divided into six steps including quantification of the source term, specification of meteorological data, atmospheric dispersion model selection, atmospheric dispersion and deposition calculation, biosphere and geosphere transport calculation, and finally dose and risk calculation. The acquisition of data and parameters for these steps were discussed in detail to provide insight to users of the integrated method. The evaluation of wind direction and speed defined in the safety analysis report of the reactor showed that the average wind direction at the reactor site altered from NE direction to NE-NNE over time but speed stayed almost the same. The maximum-total cloudshine and groundshine as a result of annual exposure rate were found as 21.1 μSv/yr at ITU Ayazağa Campus. For inhalation, Ba-140 and I-131 were found to have the highest risk in mortality and morbidity, respectively. The same is true for Sr-90 and I-131 in the lifetime intake of food while in the lifetime intake of tap water, Sr-90 is the leading radioisotope for both mortality and morbidity risks. However, all calculated values are quite low considering the survival rates. Moreover, the results of the study provides useful insight into the understanding of necessities while shedding light on limitations.

Simulation of Steam Generator Tube Rupture on the Experimental Power Reactor (RDE)

As an effort to introduce nuclear power plant in Indonesia, Indonesia’s National Atomic Energy Agency (BATAN) has been developing the small sized pebble- bed reactor of high temperature gas-cooled reactor (HTGR) technology since 2015. The 10 MWt Experimental Power Reactor (Reaktor Daya Eksperimental / RDE) was designed based on the HTR-10 design and is entering the drafting process of detail design document involving a number of calculations, experiment activities and simulations. One of research activities to verify the safety of the RDE design is by performing simulations using RELAP5 thermal hydraulic code starting from the development of RDE model for steady-state condition. One initial event classified in the safety analysis report is the steam generator tube rupture, which is simulated in this paper. The purpose of the simulation is to analyze, how accurate is the available RDE model to predict changes in the thermal behavior, particularly in the primary system. The results of two simulated cases by assuming two helical tubes ruptures, with and without safety system actuations show that the transients of pebble temperatures in the core are still under safety limit of the TRISO pebble fuel of 1230 °C. Both cases also indicate the importance of the pressure limitation devices in the primary system to control the increase of the primary pressure during the event into the allowable operating value. Other issues are still open to address such as the heat sink in the reflector model for the pebble temperature behavior, core positive reactivity due to the water-steam increase and the operation of the steam generator discharge system to mitigate the event. In general, the RELAP5 code is able to simulate the particular event, even the time sequences are different if compared with the THERMIX code.

THE VERIFICATION OF THE RSG-GAS REACTOR COOLING TOWER HEAT TRANSFER CAPACITY

The RSG-GAS reactor has been replaced and the technical specifications for the new cooling tower specify that the heat transfer capacity from the secondary cooling water to the environment is 5500 kW per module. Therefore, this study aims to verify the theoretical calculations of the heat transfer capacity using performance test data collected on the 30 MW power operation on December 20, 2018, such as the temperature of the primary and secondary coolant entering and exiting the cooling tower, wet bulb, and environmental dry bulb temperature, as well as the inlet and outlet air temperature. Furthermore, the data were used to calculate the heat transfer capacity from the secondary cooling water to the environment. The results showed that each cell of the RSG-GAS cooling tower reactor transfers the heat of approximately 5528.52 kW. This value is consistent with the technical specifications written in the revised RSG-GAS Safety Analysis Report 11.

Hemodialysis Bleeding Events and Deaths: An 18-Year Retrospective Analysis of Patient Safety and Root Cause Analysis Reports in the Veterans Health Administration.

Eighteen years of patient safety (PS) and root cause analysis reports for hemodialysis bleeding events and deaths in the Veterans Health Administration were analyzed with dual purpose: to determine the impact of a 2008 Veterans Health Administration Patient Safety Advisory on event reporting rates and to identify actions to mitigate risk and inform policy. From 2002 to 2020, 281 bleeding events (248 PS reports and 33 root cause analyses) including 14 deaths during hemodialysis treatments were identified. Events were characterized by the type of vascular access, patient mental status, and whether the access site was visible or obscured from view by staff. Of the 281 bleeding events reviewed, 188 (67%) were unwitnessed and 54 (19%) were associated with an alteration in mental status. Most deaths (n = 11; 79%) were associated with central venous catheter access. Root cause analyses reported 83 root causes, of which 33% identified physical barriers to direct observation or an equipment issue.Action plans addressed policy/procedures (30%), training/education (20%), and changes to environment/equipment (19%). Patient Safety Advisory publication was associated with a significant increase in low-harm PS reports, from 9 to 18 per year (P = 0.001). Bleeding events during hemodialysis treatments occur and may be fatal. Heightened vigilance is required when physical barriers obscure continuous direct observation, the patient exhibits an altered mental status, and vascular access is through a central venous catheter.Provider staff should consider a safety checklist and training on equipment operation. Patient Safety Advisory publication was associated with increased low-harm event reporting.

Methodology to incorporate damaged BWR spent fuel as contents in the NAC-UMS® storage system

Following Taiwan’s nuclear waste management policy, INER launched Taiwan’s first Independent Spent Fuel Storage Installation (ISFSI) project in 2005. The technology of dry storage system adopted is NAC-UMS system. INER made some design modifications of NAC-UMS system in order to accommodate the constraints at Chin-Shan Nuclear Power Plant (CSNPP). The modified UMS was named as INER High Performance System (INER-HPS). The corresponding Safety Analysis Report (SAR) of ISFSI was approved by AEC on November of 2008. Through years of dedication, the first dry run test has been accomplished on November 2012. INER-HPS is capable to cover most of the spent fuels dry storage needs in Taiwan with current NAC-UMS License. The current CoC for NAC-UMS includes undamaged BWR fuel with a burn-up of not greater than 45 GWD/MTU. For the purpose of NAC-UMS to incorporate damaged BWR fuel as approved contents, INER and NAC further enter into cooperation to developing and licensing NAC-UMS to include up to four BWR damaged fuel cans and a new BWR damaged fuel basket. This paper summarized the project progress, results of trial fabrication of damaged fuel can, and the methodology to incorporate damaged BWR spent fuel as contents in the NAC-UMS storage system.

Full scope simulation of VVER-1000 blowdown source and containment pressurization in a LBLOCA by parallel coupling of TRACE and CONTAIN

Nuclear power plants containment plays an important role as last-defined barrier in defense in depth approach against the release of radioactive material to the environment. In this study, a parallel processing couple has been developed to full scope analysis of blowdown source and containment pressurization parameters in a LBLOCA accident. To achieve this goal, primary and secondary loops of a VVER-1000/V446 were first simulated in TRACE V5.0 and steady-state results have been validated against reference data. The second step deals with containment simulation in CONTAIN 2.0 with new modified 30-cells models. A parallel processing interface was developed in MATLAB to couple TRACE and CONTAIN in the break point. Containment average pressure has been fed back to TRACE as forcing function of blowdown source in each time step during pressurization phase (coupling point). Finally, results of blowdown and containment pressurization have been validated against final safety analysis report (FSAR). Results of simulation confirm that the maximum containment pressure can reach 0.36 MPa and 0.395 MPa for this study and FSAR respectively that are lower than the maximum design absolute pressure of 0.46 MPa, so containment maintains its integrity during this accident. Temperature profiles of different control volumes inside containment during accident follow the FSAR profiles in terms of shape and value that show the ability of developed parallel coupling to full scope simulation of accidents accurately.

Small break loss of coolant accident of 200 cm² in cold leg of primary loop of Angra2 nuclear power reactor evaluation

The aim of this paper is evaluated the consequences to ANGRA 2 nuclear power reactor and to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 200cm2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. The results obtained for ANGRA 2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core.

Neutronic performance optimization of reconfigured TRIGA cores

The study deals with the neutronic analysis of the reference core and six reconfigured cores (C-1 to C-6) of the existing BAEC (Bangladesh Atomic Energy Commission) TRIGA Research Reactor using MCNP. The developed reference core model was validated by benchmarking the neutronic parameters with the experimental, calculation and Safety Analysis Report (SAR). The reconfigured cores were developed by considering three factors: (i) sub-criticality, (ii) symmetrical geometry, and (iii) substitution between fuel and graphite dummy element. Under room temperature conditions global, local and the kinetic parameters were calculated for the reconfigured cores and compared with the reference core. Cores burnup simulations were carried out up to 1200 MWD to investigate fuel depletion, flux trend, and activity accumulation. Based on neutronic performance, the results show that the reconfigured cores C-1 and C-3 are more qualified than the reference core and, in some cases; C-3 represents a better performance than the core C-1.

Transient Parameter Analysis of Non-scrammed Local Melting Accidents—A VVER 1000 Case Study

The first stage of a core degradation—based on the defense-in-depth concept of nuclear power plant (NPP) safety—is prone to fuel melting due to local blockage. The flow blockage accidents with no SCRAM happening can lead to a local fuel-clad failure, consequently affecting the safety of NPP. The present study provides an analysis of Anticipated Transient Without SCRAM (ATWS), which might lead to a condition of burning out. The accidents related to the ATWS scenarios, detailed in the case of VVER-1000/V446 reactor FSAR (Final Safety Analysis Report), include pump failure, local blockage, relative power increase, and a combination of these transients. In this research, first, drawing upon MCNPX 2.7 and COBRA-EN codes, a coupling framework is developed and then validated using an authentic reference point. The obtained results reveal that the reactor SCRAM does not occur while accidents are being investigated as there is a 10% difference in the mass flow rate reduction, a 470 kPa in the channel pressure drop, and a 204°K in the clad temperature, which constitute limitations under most pessimistic scenarios. However, under these conditions, a 70% void fraction over 12 min is observed in certain channels. Hence, burnout and local fuel melting could occur under normal operational and ATWS circumstances. According to uncertainty analyses, the occurrence of the void fraction above zero is locally definite. The transient analysis outputs could be deployed as monitoring system inputs and exploited for identifying weak points in the system.

STEADY-STATE ANALYSIS ON COOLANT SYSTEM OPERATING OF THE TRIGA2000 USING FUEL PLATE TYPE

The TRIGA2000 is one of the nuclear research reactors in Indonesia. TRIGA reactor fuel element is rod type, studies of the TRIGA reactor using plate type fuel have been carried out. One primary coolant pump is required to overcome the safety factor on the core which uses plate-type fuel. Besides that, changes in the reactor core configuration have an effect on the ability to transfer heat from the core to the coolant. Therefore, it is necessary to analyze the operating parameters of the cooling system including temperature, pressure and flow rate. This operating data is important, which is required by the operator to estimate the reactor coolant condition. The purpose of this paper is to analyze the cooling system parameters under normal steady-state operation where the primary coolant capacity is 50 kg/s and the reactor power is determined to be 1 MW. Calculation of cooling system at normal condition is carried out using the ChemCAD.6.1.4 computer program. This is also required in providing the reactor's Safety Analysis Report (SAR) document. The results show that the TRIGA reactor which uses plate-type fuel elements can be operated safely. However, in case of coolant temperature to the reactor core exceeding the design data (35 o C), it is necessary to be considered against the operational reliability of the cooling system

A Newly Developed Suppression Pool Model Based on the ISAA Code

The suppression pool is an important component in a boiling water reactor nuclear power plant. Under design-basis, loss-of-coolant accident conditions, pressure in the containment increases. Gas flows from the drywell to the wetwell after the normally submerged connecting vents between the drywell and the wetwell have been purged of water through a vent clearing process so that the suppression pool may prevent pressure in the containment from exceeding the designed pressure limit. To analyze such complicated thermal-hydraulic behavior of the suppression pool under a specific accident, an advanced reasonable model should be developed. The SPARC-90 model used in MELCOR calculates the mass and energy transfer between the bubbles of the injected gas and the suppression pool, which is affected by distance efficiency and subcooling efficiency. The dedicated vent flow model used in CONTAIN can well simulate the vent clearing time, that is, the time required for the liquid level on either side of the suppression pool to drop to the level at which the vent begins to clear. However, it is necessary and better to combine them into one integrated code. This paper presents a newly developed suppression pool model based on the self-developed severe accident analysis code Integrated Severe Accident Analysis (ISAA), which combines the advantages of the dedicated vent flow model and the SPARC-90 model to analyze the suppression pool’s thermal-hydraulic behavior. The simulation results of the developed suppression pool model shows reasonableness compared with the result in the CONTAIN 2.0 code manual. The good agreement between the simulation results and the analysis results from the COLUMBIA power station final safety analysis report demonstrates the rationality and effectiveness of the developed model, although future improvement is needed.

Analysis of loss of coolant accident without ECCS and DHRS in an integral pressurized water reactor using RELAP/SCDAPSIM

Small Modular Reactors (SMRs) are currently being developed and licensed in many countries. The innovations in the designs provide many significant features, such as decreased construction cost and time, enhanced safety, modularity, possibility of factory production etc. Several new SMR designs are based on the concept of an integral Pressurized Water Reactor. One of such designs, namely the NuScale SMR was selected for this analysis. An input model was developed from scratch for the RELAP/SCDAPSIM code using publicly available data from the NuScale Final Safety Analysis Report provided by the U.S. NRC and some validation tests have already been conducted. An extremely severe event is analyzed and discussed in the current work. It is an inside containment Loss of Coolant Accident followed by failure of both the Emergency Core Cooling System and the Decay Heat Removal System. The simulation was conducted for 86400 s (24 h) with the core damage starting after ~4.8 h, which is comparable to the results for similar scenarios presented in the references.