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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Passive Safety System Reliability Analysis: Lessons Learned and Open Items

Passive safety systems have multiple benefits, largely stemming from their high functional reliability due to their simplicity and lack of dependencies. However, accurately assessing the reliability of passive systems can be challenging due in large part to the possibility of functional failures. Argonne National Laboratory (Argonne) has been involved in several recent efforts concerning the reliability assessment of passive safety systems for advanced, non-light water reactors (non-LWRs). These efforts have focused on the development and application of mechanistic methods for the evaluation of passive system reliability, which rely on system modelling and uncertainty analyses to derive reliability values for use in probabilistic safety assessments (PSAs). Further information on these efforts is provided in ref [1]. The current paper reviews key lessons learned through these projects, along with outstanding open items regarding the reliability assessment of passive systems.

Grabaskas, David↗
DOE OSTI2016

Code Validation of a SBLOCA test in SMART Integral Test Loop

Integrated small modular reactors (iSMRs) are favored because of their unique characteristics, such as enhanced safety without any large break loss of coolant accidents (LBLOCAs), low initial construction costs, and multi-use of desalination and electricity generation. Many iSMRs have been developed thus far. In Korea, an iSMR called SMART was developed after a thorough validation of the components and systems, and finally received its design certificate in 2012 for the first time in the world. To ensure the performance of its safety systems, an integral test loop, SMART-ITL, is constructed. It contains the primary system, secondary system, passive safety systems, and an auxiliary system. The passive safety systems include a passive residual heat removal system (PRHRS), passive injection system with a core makeup tank (CMT) and a safety injection tank (SIT), and an automatic depressurization system (ADS). The height is preserved while the diameter is scaled down to 1:7. A series of SBLOCA tests have been performed with 4 trains of PRHRS, 1 train of CMT and SIT, and ADS. In parallel with the experiments, simulation of a representative test has been conducted to track the underlying thermal hydraulics. SMART-ITL is an integral test loop for SMART and a series of SBLOCA tests for confirming the performance of passive safety systems have been conducted. In this paper, a system code simulation on a SBLOCA test using SMART-ITL is presented. The reactor vessel water level, break flow rate, and CMT and SIT water levels are predicted well. (authors)

42 ENGINEERING↗
DOE OSTI2016

Experimental validation on the operational and cooling performance of the passive safety system with Atlas

After the Fukushima accident, passive safety systems are considered as the most promising alternatives to reinforce the safety and reliability of an ultimate heat removal system without any operator actions. In Korea, a passive auxiliary feedwater system (PAFS) was developed and adopted to a GEN-III+ nuclear power plant, APR+ to completely replace the conventional auxiliary feedwater system. The integral effect tests were performed to validate the cooling and operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. The representative design basis accidents (DBAs) such as a feedwater line break, a steam line break, and a steam generator tube rupture transients were experimentally simulated in this study. In addition, an integral effect test was performed for simulating a station blackout transient as a limiting beyond DBA. From the integral effect test results, it could be concluded that the APR+ has the capability of coping with the hypothetical DBA and bDBA by adopting PAFS and proper set-points of its operation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Enhanced Performance Fast Reactors with Engineered Passive Safety Systems

This report summarizes the work and results of NEUP Project #15-8251, "Enhanced Performance Fast Reactors with Engineered Passive Safety System," performed from 2015-2018 at UC Berkeley and Argonne National Laboratory. This project concerns the incorporation of a novel engineered passive safety system, called the Autonomous Reactivity Control (ARC) system, into Sodium-cooled Fast Reactors (SFRs) to improve their transient and economic performance. The first part of this project consisted of fleshing out the basic principles and the manufacturing process for an SFR fuel assembly with the ARC system installed, with the details provided in high-quality, step-by-step CAD drawings. Once the detailed design is laid out, high-fidelity simulations were performed to optimize the dimensions of important ARC system components and to quantify the standalone time-response of ARC systems in a typical SFR environment. These high-fidelity simulations were used to create reduced-order models of ARC system response which could be incorporated into the SAS4A/SASSYS-1 whole-core transient models. In order to accurately account for the time-response of ARC systems in whole-core transient simulations, modifications were made to the SAS4A/SASSYS-1 transient code. These new capabilities were tested and incorporated into the production version of the code, and have been used extensively in this analysis. Before performing whole-core transient calculations, the two reference cores selected for this study to cover a wide design space were simulated and characterized. The two reference cores, (1) a medium-sized oxide-fueled Advanced Burner Reactor and (2) a large metal-fueled Breed-and-Burn core, challenge conventional means of achieving passive safety. Because the two cores vary substantially, simulating the ARC system in both allows for conclusions to be drawn about the applicability of ARC systems in different SFRs. Once both cores were fully characterized, transient calculations were performed to understand the complicated interplay between the ARC system and other inherent reactivity feedback mechanisms. By performing parametric studies of the transient behavior, ARC system designs that best improve transient performance in each core were identified. Once transient improvements were established, the transient calculations were extended to improve reactor economics through design modifications. Upon establishing the potential benefits of ARC system inclusion, the study examines the negative neutronic implications of incorporating ARC systems into the reference cores. The study concludes by discussing a number of knowledge gaps that exist which may have important implications for ARC system implementation. Ultimately, the results of this project indicate that ARC systems may provide an effective means to improve transient performance in certain types of systems. The transient simulations in the oxide Advanced Burner Reactor showed that significant margin to coolant boiling and fuel melting could be gained through the use of a properly designed ARC system, thus improving inherent safety. While oscillatory behavior was induced in the transient response for certain combinations of ARC system parameters, the oscillation-free design space that can be effectively used to improve performance was found to be large, and the additional margin was demonstrated to allow economical improvements to the core design. Conversely, difficulties were encountered in the Breed-and-Burn core due to fundamental aspects of the core design. The combination of a particularly strong positive coolant void feedback and the use of metallic fuel allowed for diverging oscillations to manifest in response to certain transient initiators, and in the end it was found that the current ARC system design could not simultaneously enable passive safety in all transients examined. In order for complete passive safety to be achieved, a mechanism for slowing the reverse actuation of ARC systems is necessary. Regardless, the standard ARC system was able to improve the performance of two out of the three examined transients by increasing the margin to boiling in Loss of Heat Sink transients and avoiding boiling in Transient Overpower accidents. Additionally, no significant barriers to implementation were found, although the issue of tritium production should be closely monitored in any core incorporating ARC systems. Still, a number of technology gaps exist, and these are discussed at the conclusion of the report to recommend directions for future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Passive Safety System [Slides]

These slides are intended as an introduction to the concept of passive safety systems in LWRs and a discussion on designs that incorporate passive safety.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Passive Safety System Reliability Analysis: Methodology, Application, and Results

Many advanced reactor designs rely on passive systems to provide assurance that certain safety functions, such as reactivity control and reactor heat removal, are maintained during transient scenarios. The performance of these systems is sensitive to boundary and environmental conditions and can lead to functional failure, which is the inability of the system to perform as intended even without any physical failures. Additionally, conditions can cause the passive systems to operate in an intermediate or degraded state. These operational modes of passive systems and their overall reliability can be difficult to capture in traditional probabilistic frameworks where discrete operating modes are assumed or time-dependent boundary conditions are not accounted for. Novel strategies and methods are needed to be able to capture the reliability of these passive systems and their impact on facility safety. The current abstract provides a summary of such a methodology and examples of its use in a case study, the GE Hitachi (GEH) PRISM probabilistic safety assessment (PSA), and the U.S. Department of Energy (USDOE) Versatile Test Reactor (VTR) PSA.

Bucknor, M.↗
DOE OSTI2016

Systems Engineering (SE) Approach of Developing A Passive Safety Injection System (PSIS) to Cope with APR1400 Extended Station Black-Out (SBO) Accident Using RELAP5

On 11 March 2011 the power supply and cooling of the three Fukushima Daiichi reactors were disabled by a major earthquake followed by a relatively high tsunami waves. The earthquake damaged the electrical grid which is the normal power supply for the reactors, then the tsunami waves disabled the emergency diesel generators of the Fukushima Daiichi reactors, which left the plant with no alternating current (AC) source. This condition is called 'Station Black-Out'. The reactors were tripped immediately, but due to the nuclear physics phenomena, the reactor cores were still producing thermal power, which is generated by the decay of radioactive fission products in the reactor core, known as decay heat. The decay heat had not been circulated to an external heat sink and eventually the accumulative heat produced in the core led to damage to the reactor cores. After the Fukushima Daiichi accident, Design Extension Condition (DEC) such as total loss of feed water (TLOFW), and prolonged Station Black-Out (SBO) attracted the attention of nuclear experts. The attention focuses on the defense in depth concept, redundancy and sustainability of the systems, and development of the passive safety systems, which rely on natural physics laws. The topic of this paper is to discuss the possibility of attaching a Passive Safety Injection System (PSIS) to the APR1400. The PSIS is a safety system which provides adequate cooling water to the steam generators after the loss of the Auxiliary Feed Water System (AFWS) during an extended SBO accident. The PSIS will be available while the power supply of the plant is being recovered and the active safety systems take over the functionality. The anticipated PSIS will be automatically actuated to enhance the engineered safety feature and to reduce the human errors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

A multi-application small light water reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s ability to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with CFD simulation for confirmation. Furthermore, the corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Thermal-hydraulic validation tests using the SMART-ITL facility with a single train passive safety injection system for the smart design

A test program was launched using a large-scale thermal-hydraulic integral effect test facility of SMART-ITL (or FESTA) to validate the performance of the SMART passive safety system. In this paper thermal-hydraulic validation tests using the SMART-ITL facility equipped with a single train of a passive safety injection system were performed and their major results were summarized. The effect of the flow distributor, break size and different types of Core Makeup Tank (CMT) and Safety Injection Tank (SIT) on the thermal-hydraulic behavior during a Small-Break Loss of Coolant Accident (SBLOCA) scenario was investigated for the SMART passive safety system development. During the SBLOCA scenario, the water layer was well stratified thermally in the core makeup tank, and the safety injection water was injected efficiently into the reactor pressure vessel from the initial period. Eventually, the reactor coolant system was cooled down properly. From the test results, it was estimated that the SMART passive safety system has sufficient cooling capability to deal with the SBLOCA scenario of SMART. The acquired data will be used to validate the safety analysis code and its related models, to evaluate the performance of SMART PSS, and to provide the base data during the application phase of the revision of standard design approval and construction licensing. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Performance Analysis of the Passive Residual Heat Removal System of an Integral Reactor

A demand for small modular reactors (SMRs) is gradually emerging worldwide, especially in Middle East Asia and Africa. SMART (System-integrated Modular Advanced ReacTor) developed by KAERI (Korea Atomic Energy Research Institute) is the first mover among several SMRs. It is designed to produce 100 MW of electricity, or 90 MW of electricity and 40,000 tons of desalinated water concurrently, which is sufficient for 100,000 residents. SMART is currently holding the initiative as the first licensed integral reactor in the world since 4 July 2012, when the standard design approval (SDA) was officially issued by the Nuclear safety and security commission (NSSC). After that, the SMART safety enhancement project was successfully conducted with intense support of the Government, resulting in an in-depth development of the most advanced safety features including the passive residual heat removal system (PRHRS). Recently, an agreement between KAERI and King Abdullah city for atomic and renewable energy (K.A. CARE) was officially signed for the pre-project engineering (PPE) in pursuit of the construction of SMART and developing human resources capability. Both parties are currently conducting a three-year preliminary study to review the feasibility of building SMART and to prepare for its commercialization. KAERI also takes responsibility to train R and D personnel chosen from Kingdom of Saudi Arabia. Sooner or later, SMART will open the global market of SMR and be the best competitive pioneer. Since the Fukushima accident, the passive cooling system of a nuclear reactor has received a great amount of attention. SMART is designed with passive systems in order to ensure the safety of the reactor. The PRHRS is one of the passive safety systems, which is activated after an accident to remove the residual heat from the core and the sensible heat of the reactor coolant system (RCS) through the steam generators (SGs) until the condition of the RCS reaches a safe shutdown condition. In the present study, the thermal performance of the PRHRS is predicted by using in-house code. Temperature of the RCS and the cooling rate of the RCS are traced with time to see if it is satisfactorily designed to meet the design requirements of the PRHRS. The dependency of the mass flow rate of the natural circulation on the filling ratio (FR) is also discussed. The FR is arbitrarily chosen from 0.22 to 0.50 and further parametric study is needed in the future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Passive decay heat removal system design for the integral inherent safety light water reactor (I 2 S-LWR)

The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) is an innovative Pressurized Water Reactor (PWR) concept being developed by a multi-institutional team led by Georgia Tech and in collaboration with Westinghouse, under the Department of Energy’s Nuclear Energy University Programs Integrated Research Projects (DOE NEUP IRP). The University of Michigan leads the design of the thermalhydraulic and passive safety systems, in collaboration with Westinghouse and Bringham Young University. The I 2 S-LWR features an integral primary system configuration and is more conducive to the implementation of inherent safety features by eliminating potential accidents. In this paper, a novel passive Decay Heat Removal System (DHRS), is presented, consisting of a primary loop, an intermediate loop and a cooling tower loop. This passive system is designed to remove the I 2 S-LWR decay heat in the case of emergency heat removal transients, without the need for external power or operator action. The proposed DHR uses atmosphere as ultimate heat sink, to achieve indefinite decay heat removal. In this paper, firstly, the design of primary and secondary DHRS heat exchangers is optimized. Then the DHR heat removal characteristics are studied using the best-estimate thermal hydraulic code RELAP5. In addition, CFD simulations have been performed in order to investigate the DRHS helical coil primary heat exchanger performance with different coil pipe arrangements, and optimize its design. The performance of the proposed DHRS concept is investigated in case of a Station Black-Out (SBO) scenario. Operation of two, three and four DHRS trains is studied respectively. The findings show that three out of four DHRS trains are sufficient to indefinitely remove the core decay heat successfully during a SBO, and keep the reactor in a safe state without the need of any other auxiliary active system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Multiscale Experiments and Multiphysics Simulation of Multiphase Flow for Transportable Small Modular Reactors

A new type of safe, small, transportable nuclear reactor would address the intense and ever-growing global demand for energy produced via a resilient, carbon-free energy source. In this regard, transportable small modular reactors (SMRs) are being designed and developed for electricity generation within small/micro-grid/off-grid isolated systems, as well as for heat generation in industrial/residential applications. These reactors feature the capability to be fully factory fabricated and then directly transported to utilities’ sites as “plug-and-play” systems. Research and development (R&D) programs are underway at Idaho National Laboratory (INL) to successfully design, develop, and demonstrate such safe-by-design mobile reactor technologies, in collaboration with partner organizations. Multiscale experimental facilities and multiphysics simulation tools are required for reactor design verification and validation (V&V), and licensing. These advanced reactors are intended to feature passive safety systems such as passive containment cooling systems (PCCS), which consist of multiphase flows and multispecies distributions. This seminar talk will focus on designing and analyzing transportable SMR PCCS by using multiscale experiments and multiphysics computational fluid dynamics (CFD) simulations to support reactor licensing and safety. The corresponding research challenges are addressed via supportive verification and validation results generated by the models and simulation tools in combination with selective parametric and uncertainty analysis. This solution approach could blaze the trail for commercial adoption of such technologies. The facilities, simulation capabilities, and research opportunities available at INL in regard to such reactors and the integrated energy systems with which they go hand in hand are also discussed briefly, and may spark interest in deeper research as well as new collaborative projects.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Sensitivity Analyses of Natural Convection in the AP1000 Passive Containment Cooling System Following LBLOCA Using CFD

The AP1000 power plant has implemented multiple layers of passive safety systems to enhance reactor safety and ensure system integrity during postulated scenarios, such as loss of coolant accident (LOCA) and main steam line break (MSLB). Among these, the passive containment cooling system (PCCS) is a safety-related system designed to prevent the containment from exceeding the design limits of both pressure and temperature following a postulated design basis accident, by transferring heat from the steel containment vessel to the atmosphere. During LOCA, natural air convection plays a vital role in removing heat from the steel containment vessel to the atmosphere. This is integrated with the condensate film from the Passive Containment Cooling Water Storage Tank (PCCWST). Air natural convection is also the only heat removal mechanism after the dry-out of the PCCWST and during the loss of shutdown decay heat removal (LOSDHR) event. Therefore, it is essential to investigate the thermal hydraulics behavior of the containment under transient conditions to ensure its safety and integrity. This paper presents a simplified CFD/ANSYS FLUENT model developed to analyze the impact of different parameters, such as air relative humidity, air temperature, and steel containment temperature, on the natural convection capability to remove the decay heat following LOCA. The model aims to examine the thermal behavior of the containment and provides a comprehensive understanding of the system's natural convection capability during postulated accidents. The results obtained from the model can be used to improve the safety and integrity of the containment system and provide valuable insights for future research in the field.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Behavior of an air-cooled thermosiphon during a non-air gas ingress event

Here, this study investigates the transient response of an air-cooled thermosiphon subjected to a sudden non-air gas ingress event. Results presented are from experimental testing conducted on a large-scale thermal hydraulic facility that examines heat removal performance of the Reactor Cavity Cooling System (RCCS) concept, a passive safety system for advanced nuclear reactors. Relying on buoyancy driven natural circulation flow, these systems feature a series of riser standpipes and coupled chimney network that provide a pathway for heat rejection to the atmosphere. In this work, the facility was modified to allow a transition of the draft intake from normal ambient air, to draft intake from an open-top volume containing high purity argon. Testing began by establishing normal operation and allowing the facility to reach steady-state thermal hydraulic flow conditions with natural air draft. The transitioned sequence was then initiated which simultaneously changed the inlet boundary from open air to a pathway from the quiescent argon volume. This event caused 1,200 cu-feet of the heavy gas, twice the internal volume of total facility flow path, to be drawn into the inlet plenum and ingress into the heated riser standpipes. With the upper chimney network still containing residual air, there was insufficient density difference to maintain buoyancy driven natural circulation. After a period of only 90 s past the ingress event, the facility experienced complete flow stagnation before entering an extended period of severely degraded system flow. Due to the cessation of bulk fluid movement and subsequent failure of its heat removal function, fluid and structural temperatures began to rise sharply. Re-circulation patterns developed within the multiple parallel riser standpipes, where hot gas near the outlets was observed to travel downward and re-enter the inlet of adjacent channels. After approximately 18-minutes, fluid temperatures and their associated density difference rose to a level sufficient to allow reestablishment of buoyancy driven system flow, and ultimately, recovery of facility operation to normal behavior.

42 ENGINEERING↗
DOE OSTI2018

Code validation of a scaled-down DRACS model in RELAP5/SCDAPSIM/MOD 4.0

Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system proposed for Fluoride-salt-cooled High-temperature Reactors (FHRs). Additionally, benchmark study of decay heat removal capability of DRACS by natural circulation is critical for FHR safety analysis. Two scaled-down test facilities, a low-temperature DRACS test facility (LTDF) and a high-temperature DRACS test facility (HTDF), were designed following a scaling analysis and constructed to investigate thermal performance and heat removal capability of the DRACS. The LTDF uses water as a surrogate coolant for both the primary and DRACS loops. Two transient scenarios were carried out in the LTDF, namely, DRACS startup and primary coolant pump trip. In both of the scenarios, the experimental results demonstrate sufficient heat removal capabilities of the DRACS with natural circulations established in the system. These two scenarios were simulated using RELAP5/SCDASIM/MOD4.0 code and the code simulation results show good agreement with the LTDF experimental data. For the HTDF, molten salts FLiNaK and KF-ZrF4 are used respectively as the primary coolant and DRACS coolant. Thermodynamic and transport properties of FLiNaK and KF-ZrF4 were implemented into RELAP5/SCDAPSIM/MOD 4.0. Lastly, simulations of transient scenarios in the HTDF were also performed and the code results will be compared with the HTDF experimental data once available

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at the NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Molten Salt Reactor Passive Heat Removal System Modeling

The direct reactor auxiliary cooling system is a very robust, passive safety system that is designed to remove up to 2.36 MW of heat from the reactor during accident conditions. This report details a variety of DRACS degradation conditions and their effect on the safety of the reactor. This preliminary investigation shows that only two of the three DRACS loops are necessary to quickly suppress the decay heat produced by a newly shut down reactor. Even with a single DRACS loop operational, the maximum salt temperature observed was far below the safety specification of the plant (1173 K). When investigating the degraded performance of each DRACS loop, the short-term maximum salt temperature observed was strongly dependent on the DHX performance but was unaffected by the TCHX performance. However, even a heavily degraded DHX heat transfer performance was sufficient to halt the rising salt temperature due to decay heat. Further investigation should be done to characterize the effects of TCHX performance degradation at longer time scales. High levels of TCHX degradation were shown to lead to a reactor salt temperature minimum after a few hours of operation followed by a steady increase in temperature. With reduced ability to exhaust heat to the environment, it is possible the DRACS would be unable to maintain cooling during a long loss of active cooling event.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

High Fidelity Simulations of Air-Cooled Reactor Cavity Cooling System

High Temperature Gas Reactor (HTGR) designs incorporate passive safety systems (e.g., the Reactor Cavity Cooling System [RCCS]) that utilize natural principles to manage heat dissipation from the reactor pressure vessel (RPV) during accidents or routine shutdowns. The industry community is experiencing a pressing need for advanced simulation tools that can accurately assess the performance of these types of systems. In the literature, a knowledge gap exists concerning high-fidelity data for the RCCS, and this gap is one of the areas of focus of the present study. This research focuses on a specific RCCS designed for General Atomics' Modular High-Temperature Gas Reactor (GA-MHTGR). Experimental studies on a scaled version (can be seen in Figure 1) of the air-cooled RCCS used in GA-MHTGR were conducted by the University of Wisconsin-Madison (UW-Madison). This work contributes to a broader initiative aimed at establishing a numerical benchmark based on the UW-Madison experiments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗