System-integrated Modular Advanced ReacTor Research Articles

Evaluation of neutronic characteristics of accident tolerant fuel concepts in SMART reactor fuel assemblies using DRAGON

The selection of perfect accident-tolerant nuclear reactor fuels necessitates a neutronic study of the various fuels for Small Modular Reactors (SMRs). This study demonstrates a neutronic assessment of 17 × 17 fuel assemblies of Korean SMART (System-Integrated Modular Advanced ReacTor) integral type SMR using three types of Accident Tolerant Fuel (ATF) by the deterministic code DRAGON. These reactors are less expensive to build, have a shorter construction period, are safer, and have a higher power density. Uranium Nitride (UN), Uranium Di-Silicide (U3Si2), and Composite material (70 % UN- 30 % U3Si2) are considered the most potential ATF candidates. The effective multiplication factor (keff), different fissionable materials, neutron poisons, neutron-induced power, decay power, and fuel burnup are investigated as well as presented on graphs and compared from the conventional fuel UO2. Throughout the entire duration of the operation, all four models maintain a critical state while maintaining consistent dimensions. However, due to enrichment and chemical composition differences, it is found that the UN has the highest keff value. UN displayed lower concentration variation for different isotopes, whereas UO2 showed the highest power analysis study among the models. The different approaches to Accident Tolerant Fuel are not quite significant. However, opting for Uranium Nitride (UN) could offer a slight advantage in terms of effectiveness because of the higher power-up rates and burnup as well as increased cycle length when compared to the other three options.

Similarity Coefficient Generation Using Adjoint-Based Sensitivity and Uncertainty Method and Stochastic Sampling Method

In this study, a similarity coefficient generation code system was established using the Monte Carlo Code for Advanced Reactor Design (McCARD) transport code and the MIG multi-correlated input sampling code. We considered the adjoint-based sensitivity and uncertainty (S/U) and stochastic sampling (S.S.) approaches to the generation of the ck similarity coefficient. To examine the code system, the ck similarity coefficients of 23 relevant critical experiments and the System-Integrated Modular Advanced Reactor (SMART) small modular reactor (SMR) target system were generated using ENDF/B-VII.1 covariance data with the Los Alamos National Laboratory (LANL) 30-group energy group structure. Our results show that the similarity coefficients between the 16 LEU thermal-spectrum-based critical experiments and SMART are more than 0.90, which is the recommended criterion of the U.S. Nuclear Regulatory Commission (NRC). These results are very helpful for licensees and can be used to justify the determination of critical experiment benchmarks for computational bias estimations of the SMART target system. To examine the discrepancy in the similarity coefficient, ck, due to the covariance data, similarity analyses for a 24 × 24 benchmark matrix were performed using ENDF-VIII.0, JENDL-5.0, and JEFF-3.3 covariance data. The results show that the selection of the covariance data used for ck generation significantly impacts the similarity coefficient. Moreover, it was observed that the current results for the SCALE 6.1 covariance data show a consistent trend with the results reported in earlier studies.

Thermal evaluation of different integration schemes for solar-nuclear hybrid systems

Solar-nuclear hybrid systems have been proposed to meet the growing energy demand and mitigate the effects of global warming caused by fossil fuels. To achieve optimal design for these systems, an appropriate integration scheme is crucial. This article proposes three integration schemes to couple the solar system with the nuclear power plant and evaluates their performance based on energy and exergy evaluations. A reference design of the hybrid system is used, consisting of a 77.04 MWth solar tower system coupled with a 330 MWth System-Integrated Modular Advanced Reactor. Numerical results show that heating partial feedwater and saturated steam with solar energy is the best-integrated scheme, resulting in the maximum thermal efficiency and exergy efficiency for the solar-nuclear hybrid system. Comparing the standalone reactor power system to the solar-nuclear hybrid system reveals that using solar energy to heat the feedwater and saturated steam can greatly improve the energy efficiency of the power plant, increasing it to 34.97%. The solar heat to nuclear heat ratio is 25.01%, while the solar power to nuclear power ratio is as high as 44.42%, indicating effective utilization of solar heat contribution. Exergy loss distributions of the solar-nuclear hybrid system reveal that exergy losses mainly occur within the nuclear reactor and solar field, taking possession of 70% of all exergy losses.

Investigation of condensation with non-condensable gas in natural circulation loop for passive safety system

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS).There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an in-containment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance.In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

CINEMA-SMART development and its application to SBLOCA analysis for Korean small integral reactor

The Korean small integral reactor of SMART (System-integrated Modular Advanced ReacTor) was developed, which contained core, reactor coolant pumps, steam generators and pressurizer within a single reactor pressure vessel. In the code development of severe accident sequence analysis for the SMART, a CINEMA (Code for INtegrated severe accidEnt Management Analysis)-SMART has been developing from an initiation event to a containment failure including fission product behavior in the vessel and the containment. The CINEMA-SMART is composed of CSPACE, SACAP (Severe Accident Containment Analysis Package), and SIRIUS (SImulation of Radioactive nuclide Interaction Under Severe accident), which are capable of core melt progression with thermal hydraulic analysis of the RCS (Reactor Coolant System), severe accident analysis of the containment, and fission product analysis, respectively. The main sequence of a SBLOCA (Small Break Loss Of Coolant Accident) for the SMART has been analyzed using CINEMA-SMART computer code. This analysis is focused on an evaluation of core melt progression and containment thermal hydraulic behavior during a severe accident. The CINEMA results have shown that the integrity of the containment can be maintained during the SBLOCA accident in the SMART.

Machine Learning-Based Approach for Hydrogen Economic Evaluation of Small Modular Reactors

In this study, we evaluate hydrogen production costs using small modular reactors (SMRs). Furthermore, we employ a machine learning-based approach to predict important parameters that affect the hydrogen production cost. Additionally, we use a hydrogen economic evaluation program to calculate the hydrogen production cost when using the two types of SMRs: system-integrated modular advanced reactor (SMART) developed by the Korea Atomic Energy Research Institute (KAERI) and NuScale power module™ (NPM) developed by the NuScale Power, LLC. Different storage and transportation means were selected to find the cheapest option. Using SMART, storing hydrogen in compressed gas and transporting it through pipes (CG-Pipe) is the best option, with an estimated cost of USD 2.77/kg. Other options when using SMART include storing in compressed gas and transporting with a vehicle (CG-Vehicle), with an estimated cost of USD 3.27/kg; storing by liquefaction and transporting with a vehicle (L-Vehicle), with an estimated cost of USD 3.31/kg; and storing in metal hydrides and transporting with a vehicle (MH-Vehicle), with an estimated cost of USD 6.97/kg. Using NPM, CG-Pipe is the cheapest option to generate hydrogen, with an estimated cost of USD 2.95/kg. Other options include CG-Vehicle (USD 3.35/kg), L-Vehicle (USD 3.42/kg), and MH-Vehicle (USD 7.04/kg). Hydrogen production using SMART is cheaper than using NPM. However, the observed difference between the hydrogen production costs using the two reactors was insignificant. We conclude that the optimal hydrogen production cost ranges from USD 3.27/kg (CG-Vehicle) to USD 3.42 (L-Vehicle). This conclusion is because the common hydrogen transportation means is with a vehicle. From a machine learning-based approach, we determine the important parameters that affect hydrogen production costs. The most important parameter is the heat consumption (MWth/unit) at hydrogen generation plants, and other parameters include electricity rating and heat for hydrogen generation plants.

Analysis of Dynamical Marketing Strategy Modeling for Small Modular Reactor (SMR)

Using the Hubbert and System Dynamics (SD) methods, the commercialized South Korean Small Modular Reactor (SMR), the System-integrated Modular Advanced Reactor (SMART), is studied for the marking assessment using the non-linear modeling. It is shown for the calculation modeling of the Hubbert method where the period is from 2015 to 2030, because the SMR is expected to be totaled as many as 180 in 2030. However, this period is just an estimated value for this modeling. The uncertainty is expressed in the SD modeling with the described variables. In addition, several variables considered method has a different assessment trend comparing to the Hubbert equation assessment. There are some results as the SMR product, the cumulative product, and the product in maximum time. In the initial stage of the Hubbert modeling, the Product is near 2.0 which could mean the SMART export to Saudi Arabia. Some strategies to nuclear renaissance are discussed including the safety and security technology in nuclear industry with a highest priority. Furthermore, the free-human error technologies have been developed with the robotics technology.

Studi Karakteristik Termohidrolik Pada Kanal Pendingin Teras Reaktor SMART Menggunakan CFD Ansys Fluent

The Small Modular Reactor (SMR) type nuclear power plant is currently growing very rapidly by offering various kinds of safety features, both active and passive. One of the SMRs currently to be built and operated is the SMART (System-integrated Modular Advanced Reactor) reactor. The SMART reactor is designed to use conventional PWR fuel types but operates at a lower flow rate and core inlet temperature than conventional PWR. The existence of this basic difference needs to be a separate note for us in evaluating the safety of the SMART Reactor, especially for the thermohydraulic aspect. The thermohydraulic characteristics have been carried out previously using the MATRA code, however, the temperature distribution in the cooling sub channel is still unknown. Therefore, the purpose of this study is to calculate the temperature distribution in the cooling sub-channel on SMART. The program used in this study is CFD ANSYS. Ansys Fluent 2021 R1 Computational Fluid Dynamic (CFD) computer code was used to simulate coolant flow and heat transfer from fuel to coolant. The cooling sub channel was modeled consisting of four fuels with a diameter of 0.95 cm and a pitch of 1.26 cm in a rectangular arrangement surrounded by light water coolant. Based on the calculation, there is an increase in the average temperature of the coolant between the input and output sides of the coolant sub channel of 60 ℃, which is greater than the SMART reactor design. The fuel center and fuel cladding temperatures were calculated at 1100 ℃ and 350 ℃, respectively. This difference is caused by the modeling assumptions which is still far from the actual design. The model was assumed to consist of pure fuel and ignores the actual variations in fuel and fuel assembly. Keywords: SMART Reactor, CFD, Ansys Fluent, Thermal Hydraulic

Increasing cycle length and flattening power distribution in soluble boron-free small modular reactor

Population growth, industrial development as well as limited fossil fuel resources have motivated the study of other energies, especially nuclear energy. Small modular reactors have been introduced as an efficient energy source due to their greater safety, easy transport, electricity generation and water desalination, even in remote areas. Optimization in the nuclear fuel management to improve performance and save energy leads to cost-effective design with higher efficiency and better safety. In this research, core loading pattern optimization of system-integrated modular advanced reactor (SMART) has been considered using the new dragonfly algorithm. In addition to the selected algorithm, the efficiency of optimizing loading pattern also depends on the definition of the objective function. Two-objective functions including flattening the power distribution and maximizing the effective multiplication factor are considered. Simulations of the reactor fuel assemblies and reactor core were performed by DRAGON lattice calculation and PARCS core calculation codes, respectively. According to the final results, the cycle length and effective multiplication factor are increased for 185 days and 582 pcm, respectively. Also the fitness function is decreased from 0.905931 to 0.194527.

Analysis of SMART reactor core with uranium mononitride for prolonged fuel cycle using OpenMC

Abstract The neutronics performance and safety characteristics of Uranium mononitride (UN) fuel for System-Integrated Modular Advanced Reactor (SMART) has been investigated to discern the potential for non-proliferation, waste, and accident tolerance benefits of UN fuel. The neutronic evaluation of UN fuel for SMART reactor has been carried out under normal operation using OpenMC and compared with Uranium dioxide (UO2) in terms of fuel cycle length, reactivity coefficients, Fuel depletion (burnup), thermal flux, and fission product activity. The power peaking factor (PPF) has been compared at the beginning of the fuel cycle (BOC), mid of the fuel cycle (MOC), and at the end of the fuel cycle (EOC). Results indicate that the UN fuel can be operated beyond the designed fuel cycle length of the SMART reactor, which induces the positive reactivity at the end of the fuel cycle of about 4625 pcm. However, the UO2 showed negative reactivity after three years. The total fission product activity at the end of the fuel cycle (3.5 years) for UO2 and UN has been founded 1.003 × 1020 Bq and 1.023 × 1020 Bq, respectively.

Techno-Economic Assessment of Fuel Cycle Facility of System Integrated Modular Advanced Reactor (SMART)

The economic assessment of advanced nuclear power reactors is very important, specifically during the early stages of concept design. Therefore, a study was performed to calculate the total cost estimation of fuel cycle supply for a system modular advanced reactor (SMART) by using the Generation-IV economic program called G4-ECONS (Generation 4 Excel-based Calculation of Nuclear Systems). In this study, the detailed description of each model and methodology are presented including facility, operations, construction matrix, post-production model, and fuel cycle cost estimation model. Based on these models, six Generation-III+ and Generation-IV nuclear reactors were simulated, namely System 80+ with benchmark data, System 80+ with uranium oxide (UOx) and mixed oxide (MOx) fuel assemblies, fast reactor, PBMR (Pebble Bed Modular Reactor), and PWR (Pressurized Water Reactor), with partially closed and benchmarked cases. The total levelized costs of these reactors were obtained, and it was observed that PBMR showed the lowest cost. The research was extended to work on the SMART reactor to calculate the total levelized fuel cycle cost, capital cost, capital component cost, fraction of capital spent, and sine curve spent pattern. To date, no work is being reported to calculate these parameters for the SMART reactor. It was observed that SMART is the most cost-effective reactor system among other proven and advanced pressurized water-based reactor systems. The main objective of the research is to verify and validate the G4-ECONS model to be used for other innovative nuclear reactors.

Primary coolant source term of System-Integrated Modular Advanced Reactor (SMART) due to activated colloidal crud under dynamic grid load following

The grid load following result in power modulation accelerates the corrosion and activated colloidal crud activity in the primary coolant of PWRs. This paper seeks the response of activated colloidal crud in primary coolant under grid load following resulting in dynamic power modulation. Therefore, a computer code ACCP-SMART “Activated Colloidal Crud Particle in System Integration Modular and Advanced” has been developed. The core of the SMART reactor is modeled in OpenMC to generate the group fluxes under the reactor design constraints. The dynamic grid load following has been employed under the guideline and regulations of IAEA and NEA. The primary coolant activity due to activated colloidal crud has been observed as 51Cr >56Mn >99Mo >24Na >59Fe >58Co >54Mn >60Co. The activated colloidal crud with shorter half-lives (24Na, 54Mn, and 65Mn) responded swiftly corresponding to the power modulation under grid load fluctuation. However, the activated colloidal crud with longer half-lives (51Cr, 58Co, 59Fe) gradually responds to operated power maneuvering corresponding to the grid load and takes several hours (≈200 h) to attain a new saturation value after the stabilization of reactor operating power. The specific primary coolant activity due to colloidal crud is predominated in the primary coolant. Later, the steam generator activity dominated the primary coolant activity due to the gradual deposition of crud on steam generator tubes. The primary coolant specific activity due to 56Mn, 54Mn, 24Na, and 59Fe abruptly responded to reactor power modulation in primary coolant and on the core scale. while their response is trivial on steam generator surfaces. On the other hand, the primary coolant specific activity due to activated colloidal crud containing 51Cr, 58Co, and 60Co are less vulnerable and gradually respond to the power modulation and take several hours to attain a new saturation value after power stabilization.

3D-based equivalent model of SMART control rod drive mechanism using dynamic condensation method

The SMART (System-integrated Modular Advanced ReacTor) is an integral-type small modular reactor developed by KAERI (Korea Atomic Energy Research Institute). This paper discusses the feasibility and applicability of a 3D-based equivalent model using dynamic condensation method for seismic analysis of a SMART control rod drive mechanism. The equivalent model is utilized for complicated seismic analysis during the design of the SMART. While the 1D-based beam–mass equivalent model is widely used in the nuclear industry for its calculation efficiency, the 3D-based equivalent model is suggested for the seismic analysis of SMART to enhance the analysis accuracy of the 1D-based equivalent model while maintaining its analysis efficiency. To verify the suggested model, acceleration response spectra from seismic analysis based on the 3D-based equivalent model are compared to those from the 1D-based beam–mass equivalent model and experiments. The accuracy and efficiency of the dynamic condensation method are investigated by comparison to analysis results based on the conventional modeling methodology used for seismic analysis.

Development of IVR-ERVC evaluation method and its application to the SMART

An IVR (In-Vessel corium Retention) through ERVC (External Reactor Vessel Cooling) is known to be an effective means for maintaining the reactor vessel integrity during severe accidents in nuclear power plants. The evaluation method of the IVR-ERVC for small reactors was developed and applied to a small integral reactor of the SMART (System-integrated Modular Advanced ReacTor). As a first step, the thermal load analysis from the corium to the external reactor vessel wall is necessary. As a second step, the natural circulation mass flow rate analysis, which is formed in the annular gap between the external reactor vessel wall and insulator, is necessary using the thermal hydraulic computer code. The CHF (Critical Heat Flux), which is the maximum heat removal rate from the external reactor vessel wall depending on the estimated natural circulation mass flow rate, is determined using experimental results. The success criteria of the IVR-ERVC to prevent reactor vessel failure during severe accidents is determined by comparing of the thermal load with the CHF. Finally, a structure analysis on the reactor vessel wall performs to evaluate the structure integrity of the reactor vessel wall in ERVC condition using the structure analysis computer code. Consequently, the IVR-ERVC strategy turned out to be an effective means for maintaining the SMART reactor vessel integrity during severe accidents from the thermal load analysis, CHF analysis, and structure integrity analysis.

Simulation of activated corrosion product activity in Korean design system-integrated modular advanced reactor (SMART) under steady-state flow and linearly accelerated corrosion

The activated corrosion products are the leading source of radiation in Nuclear Power Plants (NPPs). Particularly in advanced integrated systems with a reduction in piping and advanced safety features such as system-integrated modular advanced reactor (SMART), the corrosion product activity is needed to be investigated. In this work, simulation of corrosion product activity has been evaluated for the SMART reactor under normal and shutdown conditions and compared with a typical Pressurized Water Reactor (PWR). For this purpose, a computer code ACP-SMART has been developed and implemented in MATLAB, which uses OpenMC as a subroutine. The effective group cross-section has been generated by OpenMC under the core design constraints. The results for 56Mn, 59Fe, 24Na, 99Mo, 58Co, and 60Co have been found as 0.072 μCi/cm3, 0.240 μCi/cm3, 0.290 μCi/cm3, 0.360 μCi/cm3, 0.015 μCi/cm3, and 0.020 μCi/cm3, respectively. The total specific activity due to corrosion products in the core, primary coolant, and inside the steam generator is 2.790 μCi/cm3, 0.183 μCi/cm3, and 0.145 μCi/cm3, respectively, which are 89.5%, 5.8%, and 4.7% of the overall contribution. Results indicate that the primary coolant and steam generator specific activity reach their saturation values fairly rapidly. Predominant corrosion product activity during normal operation is due to 56Mn while 58Co and 59Fe dominate after reactor shutdown. 99Mo remains dominant to 24Na and 59Fe during normal operation while 59Fe leads over 99Mo and 24Na after shutdown. A linearly increasing corrosion rate has been employed, and an effect on saturation activity has been investigated. For the linearly increasing corrosion rate, the specific activity behavior has changed considerably. The time taken to reach the saturation activity strongly depends upon the corrosion rate. The saturation activity in the primary coolant and steam generator depends on the CRUD removal rate from the core scale.

Study on the systematic thermal-hydraulic characteristics of helical coil once-through steam generator

Due to the compact structure, the high heat transfer efficiency and the powerful ability to withstand the stress caused by thermal expansion, Helical Coil Once-through Steam Generators (HCOTSGs) have been widely used in recent years, especially as the important equipment of Small-medium Modular Reactors (SMRs). In this study, by adopting the finite difference method, the fixed boundary method, the staggered grid to discretize the control equations and the solution regions, and using the GEAR algorithm to carry out the numerical solution, a systematic Thermal-hydraulic Analysis code for HCOTSG called TACS was developed. Compared with the design values and the calculation values of the ONCESG code of the Marine Reactor X (MRX) steam generator in Japan and the System-integrated Modular Advanced ReacTor (SMART) steam generator in Korea, the correctness of the methods and the models used in the TACS was verified. Based on the TACS, the systematic thermal–hydraulic characteristics of the HCOTSG at full load were studied with a simulation object of the steam generator of the International Reactor Innovative and Secure (IRIS). The results showed that the length of each heat transfer zone is greatly different, so heat flux changes sharply along the tube length. Secondary thermal resistances determine the change of total thermal resistance, decrease at first and suddenly rise when dry-out happens. Frictional pressure drops are the main part of secondary pressure drops, accounting for 90% of the pressure drop in the whole flow process. Besides, the transient processes in HCOTSG with the step and linear disturbances of entrance parameters were simulated by the TACS, which conformed to the analysis of the thermal–hydraulic parameters. The final stable states of these processes are the same under two disturbance modes, but the changing processes caused by linear disturbances are gentler.

Simulation of a Station Blackout Accident for the SMART Using the CINEMA Code

SMART is a small-sized integral type PWR containing major components within a single reactor pressure vessel. Advanced design features implemented into SMART have been proven or qualified through experience, testing, or analysis according to the applicable approved standards. After Fukushima accident, a rising attention is posed on the strategy to cope with a Station Blackout (SBO) accident, which is one of the representative severe accidents related to the nuclear power plants. The SBO is initiated by a loss of all offsite power with a concurrent failure of both emergency diesel generators. With no alternate current power source, most of the active safety systems that perform safety functions are not available. The purpose of SBO analysis in this paper is to show that the integrity of the containment can be maintained during a SBO accident in the SMART (System-integrated Modular Advanced ReacTor). Therefore, the accident sequence during a SBO accident was simulated using the CINEMA-SMART (Code for INtegrated severe accidEnt Management and Analysis-SMART) code to evaluate the transient scenario inside the reactor vessel after an initiating event, core heating and melting by core uncovery, relocation of debris, reactor vessel failure, discharge of molten core, and pressurization of the containment. It is shown that the integrity of the containment can be maintained during a SBO accident in the SMART reactor. It has to be mentioned that the assumptions used in this analysis are extremely conservative that the passive safety systems of PSIS and PRHRS were not credited. In addition, as ANS73 decay heat with 1.2 multiplier was used in this analysis, actual progression of the accident would be much slow and amount of hydrogen generation will be much less.

A new design concept for ocean nuclear power plants using tension leg platform

This paper presents a new design concept for ocean nuclear power plants (ONPPs) using a tension leg platform (TLP). The system-integrated modular advanced reactor, which is one of the successful small modular reactors, is mounted for demonstration. The authors define the design requirements and parameters, modularize and rearrange the nuclear and other facilities, and propose a new total general arrangement. The most fundamental level of design results for the platform and tendon system are provided, and the construction procedure and safety features are discussed. The integrated passive safety system developed for the gravity based structure-type ONPP is also available in the TLP-type ONPP with minor modifications. The safety system fully utilizes the benefits of the ocean environment, and enhances the safety features of the proposed concept. For the verification of the design concept, hydrodynamic analyses are performed using the commercial software ANSYS AQWA with the Pierson-Moskowitz and JONSWAP wave spectra that represent various ocean environments and the results are discussed.

Safety analysis of pool‐type double containment of system‐integrated modular advanced reactor: A case study for Saudi Arabia

In March 2015, the Kingdom of Saudi Arabia signed an agreement with South Korea to build the system-integrated modular advanced reactor (SMART) originally developed by the Korean Atomic Energy Research Institute. The SMART is a structurally contained safe dual-purpose power plant reactor that produces both electricity and desalination. This study aims to design a double-containment structure with a pool-type concept to enhance the SMART safety and develop a robust design for future modular reactors. For that purpose, a theoretical model is developed herein, which leads to the modeling of a double-containment structural design of a water pool. The model calculates various parameters (ie, temperature, pressure, phasic velocities, and condensable gases) and is used to develop a simulation program for use in a computational assessment. Moreover, two best-estimate thermal hydraulic engineering programs, namely THEATRe and RELAP5, are used to simulate the original SMART with the double-containment water pool structure. The differences in the modeling strategies of the two simulation techniques are discussed to accurately measure various technical parameters. Accordingly, a nodalization diagram of the reactor and the pool-type concept are developed and simulated. The results obtained from the computational models yield the error of >2%, depicting accuracy. A small-break loss of coolant accident is also simulated to mitigate the transient conditions. The result verifies the adoptability of this reactor concept in any other small modular reactor.

Comparison of Two Different Sized Small-Break LOCAs on the Passive Safety Injection Line Using SMART-ITL Data

Even for small modular reactors (SMRs) with all large pipes removed, a small-break loss-of-coolant accident (SBLOCA) is an important design-basis accident (DBA). Experimental simulation of the SBLOCA scenario is essential before a prototype reactor is realized. The system-integrated modular advanced reactor (SMART) is one of the SMRs developed by the Korea Atomic Energy Research Institute. An integral test loop, SMART-ITL, was also constructed to carry out several types of integral thermal-hydraulic effects tests for the prototype reactor. The SMART-ITL was designed with a preserved height, 1/7th diameter, and 1/49th area, and volume-scaling ratios. Two types of passive safety systems were equipped in the SMART-ITL: a passive safety injection system (PSIS) and a passive residual heat removal system (PRHRS). The PSIS was designed to refill the coolant in the reactor coolant system (RCS) for 72 h after an accident. Under accident conditions the PRHRS prevents overheating and overpressurization of the RCS using two-phase natural circulation. The SBLOCA on the passive safety injection line is a significant DBA that should be validated for differences in break size. In this paper, the effects of two different break sizes, 2 and 7/32 in., were analyzed in order to study the effect of the maximum and minimum mass and energy loss of the RCS. In order to simulate a clear difference between maximum and minimum mass and energy loss of the RCS, heat removal by the PRHRS was performed in the maximum break size (2-in.) accident, and heat removal by the PRHRS was not conducted in the minimum break size (7/32-in.) accident. The difference in mass and energy loss of the RCS will have a significant impact on the operation of the automatic depressurization system. Using the two extreme accident simulations, it was possible to confirm the difference in accident progression caused by the difference in break size and the characteristics of the PSIS.