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The four different geometries of a reactor vessel auxiliary cooling system (RVACS) are designed to investigate how the heat removal capability changes by the geometries. Each geometry has different heat transfer characteristics regarding airflow and the presence of an air separator and an insulation material. The heat removal performance is evaluated with the reactor vessel (RV) temperature, the RV wall emissivity, and the airflow gap. For the same RVACS volume, 600 °C RV temperature, and 6-cm gap, the heat removal capability varies from 240.8 kW to 308.5 kW, depending on the geometry. The wall emissivity is less effective for Geometry 2, which has a large cavity volume and a small heat transfer area compared to the other geometries. The highest heat removal performance was obtained using Geometry 3 because cold air flows in from the bottom of the RVACS, improving both radiative and convective heat transfer. Reducing the gap size by 3 cm results in only 80.0 kW of heat removal capability for Geometry 1, and the heat removal dramatically decreases to 0.2 kW at an RV temperature of 800 °C. A sufficient RVACS gap size of at least 6 cm with a 0.6-m diameter intake pipe is required to provide adequate natural circulation and eventually enhance the heat removal capability.

In the event of a core meltdown in a high-power reactor, the integrity of the reactor pressure vessel is presumably protected by severe accident mitigation systems such as in-vessel retention external reactor vessel cooling (IVR-ERVC). However, in the late phase of the accident, two possible locations on the RPV are prone to failure: the location of the focusing effect and location of in-core instrument penetration. These two potential points of damage in the RPV are investigated in this study. A numerical model for the prediction of the natural convection, melting, and solidification processes for IVR-ERVC is presented. The model is based on the enthalpy-porosity approach with an extension for continuous liquid fraction function. The model is implemented in open-source field operation and manipulation (OpenFOAM) computational fluid dynamic code to produce a new solver which is based on the combination of conjugate heat transfer solver and buoyant-driven natural convection solver and the new solver is validated against the melting Gallium experimental test, in-core instrumentation failure experimental test, and BALI experimental test. This numerical model is applied for the investigation of the RPV rupture at the location of the focusing effect and in-core instrumentation penetrations. Severe ablations of the cladding and the weld materials are observed at a heat load of about ~1800 K which is expected to lead to the ejection of the penetration tubes if the force holding the penetration tube in place is lower than the force exerted by the system pressure. Subsequently, a two-layer IVR configuration is assessed and the integrity of the RPV is found not to be compromised under external reactor vessel cooling. However, in the case of a boiling crisis, the temperature of the ex-vessel wall is expected to rise quickly and this is simulated by increasing the ex-vessel wall temperature. The RPV is found to fail near the beltline due to a phenomenon known as focusing effect when the ex-vessel wall temperature rises above 1200 K.

In a sodium-cooled fast reactor, a sodium-water reaction (SWR) could be occurred at pressure boundaries between the two separated systems, the steam Rankine cycle and the sodium coolant system, which results in a serious problem in the system. Since it has been a design challenge for researchers, as an alternative system, the supercritical CO 2 (S-CO 2 ) Brayton cycle was suggested. To utilize the S-CO 2 Brayton cycle to a SFR, a series of studies on the coupled system have been conducted at Korea Atomic Energy Research Institute (KAERI). The studies were mainly focused on the pressure boundary failure events and their consequences including Na-CO 2 reactions. Prior to the experimental studies, the authors classified the reaction behaviors of Na-CO 2 reaction based on the possible scenarios that could take place at pressure boundaries. Through experimental studies, it was found that the reaction follows typical chemical formulas and it is essentially dependent on the reaction temperature. More importantly, the reaction rate changed drastically at the specific temperatures and the reaction became flammable exceeding a certain temperature. To specify the design conditions of the SFR system coupled with the S-CO 2 Brayton cycle, by means of a dedicated set-up the self-ignition temperature has been evaluated. Exceeding this threshold temperature a significant increase in the reaction speed is observed. We setup an experimental apparatus and a test matrix based on one of possible scenarios where the flammable Na-CO 2 reaction takes place and we installed a visualized chemical reactor to watch reaction phenomenon around the self-ignition temperature. The experimental apparatus was designed to be easy to control heaters, to regulate gas supply, and to measure reaction temperature. Through the experiment, it was found that the flammable temperature of Na-CO 2 reaction lies between 598.3˚C and 599.9˚C and how liquid sodium reacts with gaseous carbon dioxide at these temperatures. The results will be help to set design criteria for the pressure boundary and it will take long time but an appropriate safety system would be introduced.

A parametric study that demonstrates a methodology for determining the optimum bilayer composition in a duplex SiC cladding is discussed. The structural performance of multi-layer SiC cladding design is significantly affected by radial thickness fraction of each layer. This study shows that there exists an optimal composite/monolith radial thickness fraction that minimizes failure probability for a duplex SiC cladding in steady-state operation. An exemplary reference case study shows that the duplex cladding with the inner composite fraction ∼0.4 and the outer CVD-SiC fraction ∼0.6 is found to be the optimal SiC cladding design for the current PWRs with the reference material choice for CVD-SiC and fiber reinforced composite. A marginal increase in the composite fraction from the presented optimal designs may lead to increase structural integrity by introducing some unquantified merits such as increasing damage tolerance. The major factors that affect the optimum cladding designs are temperature gradients and internal gas pressure. Clad wall thickness, thermal conductivity, and Weibull modulus are among the key design parameters/material properties.

Hybrid micro modular reactor (H-MMR) combining MMR with renewable energy and energy storage systems (ESS) is designed using heat pipe. H-MMR could produce 10 MW of electric power more flexibly and efficiently through load following. LOCA accident is precluded in advance as there is no primary coolant due to the use of potassium heat pipe. Reactor solid core consist of hexa-annulus UN fuel with a heat pipe inserted, which connects a sodium pool of intermediate heat exchanger (IHX). The annular gap wick heat pipe (ACHP) using potassium is optimized to maximize the heat transport performance. Reactor core configuration is designed through core analysis and thermal analysis using 1D lumped finite differential method. Reactor core of H-MMR make 18 MW of thermal power for 56 years without refueling when diameter of heat pipe is 22 mm. The feasibility of heat pipe safety system and sodium pool is evaluated in normal operation. Reactor vessel auxiliary cooling system (RVACS) is designed to cool down decay heat within safety margin using implicit transient analysis.

This study presents the importance of coherency in modeling thermal-hydraulics and mechanical behavior of a solid for an advanced prediction of cladding thermal shock fracture. In water quenching, a solid experiences dynamic heat transfer rate evolutions with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates has been overlooked in the analysis of thermal shock fracture. In this study, we are presenting quantitative evidence against the prevailing use of a constant heat transfer coefficient for thermal shock fracture analysis in water. We conclude that no single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials. The presented results show a remarkable stress prediction improvement up to 80–90% of the actual stress with the use of the surface temperature dependent heat transfer coefficient. For thermal shock fracture analysis of brittle fuel cladding such as oxidized zirconium-based alloy or silicon carbide during LWR reflood, transient subchannel heat transfer coefficients obtained from a thermal-hydraulics code should be used as input for stress analysis. Such efforts will lead to a fundamental improvement in thermal shock fracture predictability over the current experimental empiricism for cladding fracture analysis during reflood.

One of the promising future nuclear energy systems, the Sodium-cooled Fast Reactor (SFR) has been actively developed internationally. Recently, to improve safety and economics of a SFR further, coupling supercritical CO2 power cycle was suggested. However, there can be a chemical reaction between sodium and CO2 at high temperature (more than 400 °C) when the pressure boundary fails in a sodium–CO2 heat exchanger. To ensure the performance of such a system, it is important to employ a cleaning agent to recover the system back to normal condition after the reaction. When sodium and CO2 react, solid and gaseous reaction products such as sodium carbonate (Na2CO3) and carbon monoxide (CO) appear. Since most of solid reaction products are hard and can deteriorate system performance, quick removal of solid reaction products becomes very important for economic performance of the system. Thus, the authors propose the conceptual method to remove the byproducts with a chemical reaction at high temperature. The chemical reaction will take place between the reaction byproducts and a cleaning agent while the cleaning agent is inert with sodium. Thus, various sodium-based compounds were first investigated and three candidate substances satisfying several criteria were selected; sodium bromate (NaBrO3), sodium chlorate (NaClO3), and sodium tetrafluoroborate (NaBF4). The selected substances were thermally analyzed with the TG/DTA studies. Unfortunately, it was revealed that all candidate substances did not react with Na2CO3 and decomposed before 600 °C. However, since no study has been performed on the issue of cleaning byproducts of Na–CO2 reaction so far, this study provides the basic guideline for the future study and suggests the future research direction and preferred characteristics of cleaning agent of sodium CO2 reaction byproducts.

In nuclear safety analyses, a 1-D thermal hydraulic system analysis code with the best estimate plus uncertainty (BEPU) approach is generally used. There are sources of uncertainty such as nodalization, physical models, and numerical scheme. Among them, the authors specifically focused on the node configuration uncertainty related to spatial discretization that is regarded as less dominant sources of uncertainty in the code previously. Furthermore, it was not easy to quantify the uncertainty from the node configuration. The authors quantified and analyzed the node configuration uncertainty to understand how much it affects the solution. In this study, the effect of spatial discretization are studied by discussing simulation results for various spatially discretized node systems. The discussion on the node configuration uncertainty will be presented by examples of SUbcooled BOiling (SUBO) test simulation and Loss Of Fluid Test (LOFT) simulation using Multi-dimensional Analysis of Reactor Safety (MARS) code developed in South Korea. From SUBO test simulation, it was found that when the number of nodes is small than the node configuration uncertainty can overwhelm the uncertainty from the physical model. From LOFT simulation, it was revealed that the node configuration can cause bifurcation in the complex nuclear system analyses. Thus, it is concluded that the node configuration uncertainty can be quite significant and has to be considered in the nuclear system analyses.

With the increased role of intermittent renewables in the energy mix, it is important for the Nuclear Power Plants (NPPs) to develop flexible load-follow capabilities in order to be economically competitive. By combining NPP with external superheaters, there is possibility to increase power conversion cycle efficiency and reduce load to some extent while maintaining the nuclear reactor operation at 100% of its rated power and thus maximizing its economic value. In this paper, parametric study of a Small Modular Boiling Water Reactor (SMBWR) conceptual design combined with external superheaters is presented quantifying its performance at different system pressure. The first part of the work is focusing on how operating pressure will affect the SMBWR system as a whole, which includes neutronics, thermal hydraulics, and thermodynamics. WIMS-PANTHER package and COBRA-EN were used as the neutronic and thermal hydraulic tools. In addition, MATLAB models of the SMBWR natural circulation loop and balance of plant were developed for the study. The second part of the work examines the maneuvering capability of SMBWR. It is found that increasing the pressure from 6.5 to 10.0 MPa has no significant neutronic effects, while thermal efficiency is slightly improved. It is also found that SMBWR is able to reduce the load to 65% while maintaining the reactor power at 100% of its rated value.

From the Fukushima accident, protection of NPPs from any imaginable natural disasters became very important. In this study, the authors suggest a new concept of ocean nuclear power plant (ONPP) by using SMART as a reference reactor, which is the most recent Small Modular Reactor (SMR) developed by Korea, to demonstrate that the proposed concept can improve the safety of NPP from earthquake and tsunami. The proposed concept utilizes Gravity Based Structure (GBS), which is a widely spread construction technique of offshore plants. Because, floating type or submerged type NPPs can be easily affected by severe ocean environments such as tsunamis and storms, additional safety features have to be added to the existing land based plant. In contrast, the newly proposed GBS-type ONPP does not require going through significant design modifications due to inherent characteristics of the construction method. The authors have demonstrated this concept can be applied to the large nuclear power plant in the previous work and will expand this concept for SMRs in this paper. The authors discuss the new concept by presenting design parameters, design requirements, and the new total general arrangement. Furthermore, due to the unique configuration of ONPP SMART, innovative passive safety features can be added to the existing SMART design. The performance of proposed concept to resist earthquake as well as newly added passive safety feature will be discussed by presenting simplified analysis results.

Many previous sodium cooled fast reactors (SFRs) adopted steam Rankine cycle as the power conversion system. However, the concern of sodium water reaction has been one of the major design issues of a SFR system. As an alternative to the steam Rankine cycle, several closed Brayton cycles including supercritical CO2 cycle, helium cycle and nitrogen cycle have been suggested recently. In this paper, these alternative gas Brayton cycles will be compared to each other in terms of cycle performance and physical size for small modular SFR application. Several new layouts are suggested for each fluid while considering the turbomachinery design and the total system volume.

The Supercritical Carbon Dioxide (S-CO2) Brayton cycle has been gaining attention due to its compactness and high efficiency at moderate turbine inlet temperature. Previous S-CO2 cycle research works in the field of nuclear engineering were focused on its application to the next generation reactor with higher turbine inlet temperature than the existing conventional water-cooled nuclear power plants. However, it was shown in authors’ previous paper that the advantages of the S-CO2 Brayton cycle can be also further applied to the water-cooled Small Modular Reactor (SMR) with a success, since SMR requires minimal overall footprint while retaining high performance. One of the major issues in the S-CO2 Brayton cycle is the selection and design of appropriate turbomachinery for the designed cycle. Because most of the nuclear industry uses incompressible working fluids or ideal gases in the turbomachinery, a more detailed examination of the design of the turbomachinery is required for a power system that uses S-CO2 as working fluid. This is because the S-CO2 Brayton cycle high efficiency is the result of the non-ideal variation of properties near the CO2 critical point. Thus, the major focus of this paper is to suggest the design of the turbomachinery necessary for the S-CO2 Brayton cycle coupled to water cooled SMRs. For this reason, a S-CO2 Brayton cycle turbomachinery design methodology was suggested and the suggested design methodology was first tested with the existing experimental data to verify its capability. After then, it was applied to the proposed reference system to demonstrate its capability and to provide fundamental information for the future design.

Passive safety is a primary motive behind the development of small and medium sized reactors of various coolants. After the Fukushima accident, there is an increased interest in a nuclear reactor's reliance on passive safety systems. Most of the existing passive systems, regardless of the reactor type, utilize buoyancy force to drive the cooling flow. Hence, it is essential to evaluate if the naturally developed cooling flow is sufficient to maintain the heated surface temperature of the fuel elements below the design limit. Evaluating passively driven flows can be quite a challenging task in both two phase natural circulation systems and also in single phase natural circulation systems. Previous research works have found that single phase heat transfer can be deteriorated and becomes uncertain when the driving force of a system is shifted from external force (forced convection) to self generated buoyancy force or a combination of both (natural or mixed convection). In this paper, single phase gas, water, and liquid metal reactors with passive systems are reviewed briefly. A simple theoretical analysis of each reactor type is performed to find the tendency of the shift in the operating heat transfer regime into the deteriorated region. The analysis results show that single phase water system can maintain operation within the forced convection regime but the operating regime gets closer to the deteriorating heat transfer regime as the system's physical size reduces from a large nuclear power plant to the small and medium reactor scale. The gas cooled system has a high tendency to operate in the deteriorated heat transfer regime when the driving force changes from forced to natural. Meanwhile the liquid metal system demonstrates more margins to operate outside from the deteriorated heat transfer region compared to the two other fluid systems. However further studies are needed to clearly identify the boundaries of the deteriorated heat transfer regime for each coolant since the deterioration greatly depends on the thermophysical properties variation of the coolant and the near-wall flow behavior of the coolant with respect to temperature change.

In this paper, we present a new concept for offshore nuclear power plants (ONPP) with enhanced safety features. The design concept of a nuclear power plant (NPP) mounted on gravity-based structures (GBSs), which are widely used offshore structures, is proposed first. To demonstrate the feasibility of the concept, a large-scale land-based nuclear power plant model APR1400, which is the most recent NPP model in the Republic of Korea, is mounted on a GBS while minimizing modification to the original features of APR1400. A new total general arrangement (GA) and basic design principles are proposed and can be directly applied to any existing land based large scale NPPs. The proposed concept will enhance the safety of a NPP due to several aspects. A new emergency passive containment cooling system (EPCCS) and emergency passive reactor-vessel cooling system (EPRVCS) are proposed; their features of using seawater as coolant and safety features against earthquakes, Tsunamis, storms, and marine collisions are also described. We believe that the proposed offshore nuclear power plant is more robust than conventional land-based nuclear power plants and it has strong potential to provide great opportunities in nuclear power industries by decoupling the site of construction and that of installation.

The supercritical carbon dioxide (S-CO2) Brayton cycle is being considered as a favorable candidate for the next generation nuclear reactors power conversion systems. Major benefits of the S-CO2 Brayton cycle compared to other Brayton cycles are: (1) high thermal efficiency in relatively low turbine inlet temperature, (2) compactness of the turbomachineries and heat exchangers and (3) simpler cycle layout at an equivalent or superior thermal efficiency. However, these benefits can be still utilized even in the water-cooled reactor technologies under special circumstances. A small and medium size water-cooled nuclear reactor (SMR) has been gaining interest due to its wide range of application such as electricity generation, seawater desalination, district heating and propulsion. Another key advantage of a SMR is that it can be transported from one place to another mostly by maritime transport due to its small size, and sometimes even through a railway system. Therefore, the combination of a S-CO2 Brayton cycle with a SMR can reinforce any advantages coming from its small size if the S-CO2 Brayton cycle has much smaller size components, and simpler cycle layout compared to the currently considered steam Rankine cycle. In this paper, SMART (System-integrated Modular Advanced ReacTor), a 330 MWth integral reactor developed by KAERI (Korea Atomic Energy Institute) for multipurpose utilization, is considered as a potential candidate for applying the S-CO2 Brayton cycle and advantages and disadvantages of the proposed system will be discussed in detail. In consideration of SMART condition, the turbine inlet pressure and size of heat exchangers are analyzed by using in-house code developed by KAIST–Khalifa University joint research team. According to the cycle evaluation, the maximum cycle efficiency under 310 °C is 30.05% at 22 MPa of the compressor outlet pressure and 36% of flow split ratio (FSR) with 82 m3 of total heat exchanger volume while the upper bound of the total cycle efficiency is 37% with ideal components within 310 °C. The total volume of turbomachinery which can afford 330 MWth of SMR is less than 1.4 m3 without casing. All the obtained results are compared to the existing SMART system along with its implication to other existing or conceptual SMRs in terms of overall performance in detail.

A sodium-cooled fast reactor (SFR) is one of the strongest candidates for the next generation nuclear reactor. However, the conventional design of a SFR concept with an indirect Rankine cycle is subjected to a possible sodium–water reaction. To prevent any hazards from sodium–water reaction, a SFR with the Brayton cycle using Supercritical Carbon dioxide (S-CO2) as the working fluid can be an alternative approach to improve the current SFR design. However, the S-CO2 Brayton cycle is more sensitive to the critical point of working fluids than other Brayton cycles. This is because compressor work is significantly decreased slightly above the critical point due to high density of CO2 near the boundary between the supercritical state and the subcritical state. For this reason, the minimum temperature and pressure of cycle are just above the CO2 critical point. In other words, the critical point acts as a limitation of the lowest operating condition of the cycle. In general, lowering the rejection temperature of a thermodynamic cycle can increase the efficiency. Therefore, changing the critical point of CO2 can result in an improvement of the total cycle efficiency with the same cycle layout. A small amount of other gases can be added in order to change the critical point of CO2. The direction and range of the critical point variation of CO2 depends on the mixed component and its amount. Several gases that show chemical stability with sodium within the interested range of cycle operating condition were chosen as candidates for the mixture; CO2 was mixed with N2, O2, He, and Ar. To evaluate the effect of shifting the critical point and changes in the properties of the S-CO2 Brayton cycle, a supercritical Brayton cycle analysis code with a properties program, which has the most accurate mixture models, was developed. The CO2–He binary mixture shows the highest cycle efficiency increase. Unlike the CO2–He binary mixture, the cycle efficiencies of CO2–Ar, CO2–N2, and CO2–O2 binary mixtures decreased compared to the pure S-CO2 cycle. It was found that the increment of critical pressure led to a decrease in cycle operating pressure ratio which resulted in a negative effect on total cycle efficiency. In addition, the effects from changed minimum operating condition and property variations of multi-component working fluid changed the recuperated heat in the cycle which was closely related to the cycle performances.

Sulfur–iodine (SI) cycle should overcome many engineering challenges to commercialize and prove its feasibilities to compete other thermo-chemical cycles. Some critical issues such as structural material, harsh operating condition and high capital costs were considered obstacles to be actualized. Operating SI cycle at low-pressure is one of the solutions to actualize the cycle. The flash operation with over-azeotropic HI at low pressure does not require temperature and pressure as high as those in the existing methods as well as heating for separation. The operation in low pressure reduces corrosion problems and enables us to use flexible selection of structural material. We devised an intermediate heat exchanging–depressurizing loop to eliminate high operating pressure in the hydrogen side as well as a large pressure difference between the reactor side and the hydrogen side. Molten salts are adequate candidates as working fluids under the high-temperature condition with homogeneous phase during pressure changing process. Using molten salts, 2.20–4.65 MW of pumping work is required to change the pressure from 1 bar to 7 MPa. We selected BeF2-containing salts as the possible candidates based on preliminary economic and thermal hydraulic consideration.

Steady-state natural circulation data obtained in a 7 m-tall experimental loop with carbon dioxide and nitrogen are presented in this paper. The loop was originally designed to encompass operating range of a prototype gas-cooled fast reactor passive decay heat removal system, but the results and conclusions are applicable to any natural circulation loop operating in regimes having buoyancy and acceleration parameters within the ranges validated in this loop. Natural circulation steady-state data are compared to numerical predictions by two system analysis codes: GAMMA and RELAP5-3D. GAMMA is a computational tool for predicting various transients which can potentially occur in a gas-cooled reactor. The code has a capability of analyzing multi-dimensional multi-component mixtures and includes models for friction, heat transfer, chemical reaction, and multi-component molecular diffusion. Natural circulation data with two gases show that the loop operates in the deteriorated turbulent heat transfer (DTHT) regime which exhibits substantially reduced heat transfer coefficients compared to the forced turbulent flow. The GAMMA code with an original heat transfer package predicted conservative results in terms of peak wall temperature. However, the estimated peak location did not successfully match the data. Even though GAMMA's original heat transfer package included mixed-convection regime, which is a part of the DTHT regime, the results showed that the original heat transfer package could not reproduce the data with sufficient accuracy. After implementing a recently developed correlation and corresponding heat transfer regime map into GAMMA to cover the whole range of the DTHT regime, we obtained better agreement with the data. RELAP5-3D results are discussed in parallel.

The thermal-hydraulic performance of the PCHE was investigated using the KAIST helium test loop. Experiments were performed in the helium laminar region with 350

Thermal striping, characterized by turbulent mixing of two flow streams of different temperatures that result in temperature fluctuations of coolant near the pipe wall, is one of the main causes of thermal fatigue failure. Coolant temperature oscillations due to thermal striping are on the order of several Hz. Thermal striping high-cycle thermal fatigue that occurs at tee junctions is one of the topics that should be addressed for the life management of light water reactor (LWR) piping systems. This study focuses on numerical analyses of the temperature fluctuations and structural response of coolant piping at a mixing tee. The coolant temperature fluctuations are obtained from Large Eddy Simulations that are validated by experimental data. For the thermal stress fatigue analysis, a model is developed to identify the relative importance of various parameters affecting fatigue-cracking failure. This study shows that the temperature difference between the hot and cold fluids of a tee junction and the enhanced heat transfer coefficient due to turbulent mixing are the dominant factors of thermal fatigue failure of a tee junction.

Transient response of a Gas Cooled Fast Reactor (GFR) coupled to a recompression supercritical CO2 (S-CO2) power conversion system (PCS) in a direct cycle to a Loss of Coolant Accident (LOCA) and a Loss of Generator Load Accident is analyzed using RELAP5-3D. A number of thermal hydraulic challenges for GFR design are pointed out as the designers strive to accommodate cooling of the high power density core of a fast reactor by a gas with its inherently low heat transfer capability, in particular under post-LOCA events when system pressure is lost and when reliance on passive decay heat removal (DHR) is emphasized. Although it is possible to design a S-CO2 cooled GFR that can survive LOCA by cooling the core through natural circulating loops between the core and elevated emergency cooling heat exchangers, it is not an attractive approach because of various bypass paths that can, depending on break location, degrade core cooling. Moreover, natural circulation gas loops can operate in deteriorated heat transfer regimes with substantial reduction of heat transfer coefficient: as low as 30% of forced convection values, and data and correlations in these regimes carry large uncertainties. Therefore, reliable battery powered blowers for post-LOCA decay heat removal that provide flow in well defined regimes with low uncertainty, and can be easily overdesigned to accommodate bypass flows were selected. The results confirm that a GFR with such a DHR system and negative coolant void worth can withstand LOCA with and without scram as well as loss of electrical load without exceeding core temperature and turbomachinery overspeed limits.