Reactor Cavity Cooling System Design Research Articles

Modeling and validation of a water-cooled reactor Cavity cooling system (RCCS)

High Temperature Gas-cooled Reactors (HTGRs) have garnered interest for use in commercial energy production and for providing high temperature process heat for industrial applications. One of the passive safety systems suggested for decay heat removal in accident conditions is known as the Reactor Cavity Cooling System (RCCS). Two steady-state experimental data sets with varying Reynolds numbers (Re = 2490, 10279) conducted with the TAMU RCCS experimental facility were modeled and simulated using the computational fluid dynamics code STAR-CCM + and the TRACE system thermal hydraulics code for code validation. It was found that both the CFD and system code models showed good agreement with respect to the global conditions of the internal flow within the RCCS. Comparisons of the local temperature measurements showed that the TRACE code tended to overpredict temperatures toward the bottom of the RCCS cooling panel but showed good agreement above the midpoint of the RCCS risers. The CFD model found significant thermal stratification and recirculation in the Low Reynolds case due to the asymmetric heating profile applied to the riser panel. The bottom and lower manifold geometries of the RCCS design were also found to produce asymmetrical flow conditions which led to differences in local riser temperatures.

Design of a prototypical natural circulation water-based reactor cavity cooling system (RCCS) for a pebble-bed generic FHR

A prototypical natural circulation water-based reactor cavity cooling system (RCCS) for the UC Berkeley Mark-1 pebble-bed gFHR is designed based on one-dimensional thermal hydraulics modeling and optimization implemented in an in-house MATLAB code. The model employs a lumped core consisting of fuel pebble and graphite pebble regions, while the graphite reflector, vessel, and RCCS are represented as separate but energy-coupled regions. The model is derived based on steady state energy balance equations accounting for conductive heat transfer from the lumped core to the vessel and subsequent radiative heat transfer to the RCCS and convective heat transfer in the water. Mass flow rate is calculated based on momentum and thermal energy balance in the RCCS. Starting with ANL’s water NSTF as baseline, effects of RCCS design parameters such as size of the plate between pipes, pipe diameter, source-to-sink distance, surface emissivity, and pitch from the core are examined in an integral effects framework which accounts for subsequent variation of the mass flow rate, number of pipes, convection coefficient, mean region and surface temperatures, radiative view factors, and water outlet temperature. The importance of including a prototypical reactor design in RCCS design calculations is emphasized as the calculations show that temperature drops from the core to the RCCS fluid to transfer the heat tend to be the limitation not the heat removal capacity of the fluid. A closed-loop design is obtained with capability to safely remove up to 0.72% of the nominal reactor power at maximum estimated peak conditions and 0.40% at shutdown. The design is based on physics calculations and does not account for economic optimization.

Sensitivity of SC-HTGR Conduction Cooldown to Reactor Cavity Cooling System Failure

The inherent passive heat removal characteristics of modular High Temperature Gas-Cooled Reactors (HTGRs) are well known. Modular HTGRs use a combination of coated-particle fuel, ceramic core materials, core geometry, and power level to maintain acceptable fuel temperatures for all credible operating and accident conditions. Heat from the reactor vessel is radiated to a passive reactor cavity cooling system (RCCS), which removes excess heat from the reactor cavity. The RCCS for Framatome’s Steam Cycle–High Temperature Gas-Cooled Reactor (SC-HTGR) is a highly reliable, redundant system. Similar to most other modular HTGR concepts, RCCS failure is not considered credible for any accident scenario. Nonetheless, reactor module performance with a compromised RCCS is still of interest. Evaluation of such beyond-design-basis scenarios supports safety assessment of extremely low probability beyond-design-basis events (BDBEs) as well as the development of RCCS design requirements and plant emergency procedures. This study evaluates the performance of the SC-HTGR during a long-term depressurized loss of forced circulation event without RCCS operation. Boundary conditions are varied to determine their effect on reactor temperatures. Safety and investment risk considerations are addressed. The results of this study indicate that the safety impact is modest since fuel temperatures remain within their limits. However, the investment risk is more significant since vessel temperatures could significantly exceed design limits for these hypothetical BDBEs.

Comparative study with two reduced test facilities for air-cooled RCCS scaling law

Korea Atomic Energy Research Institute (KAERI) and Argonne National Laboratory (ANL) conducted steady-state tests at constant heat flux conditions to investigate the scale effect on heat removal behavior of the air-cooled Reactor Cavity Cooling System (RCCS). Two differently scaled-down test facilities were used in this study, KAERI’s ¼ scale with 4.0-m cavity height, and ANL’s ½ scale with 6.8-m cavity height. Two scaling laws were proposed to simulate the radiation across the cavity and the buoyancy-driven duct flow in the riser, respectively. The test matrix for KAERI and ANL test facilities focused on the scaling effect of the heated riser length with constant heat flux at the PMR200 RCCS design. The test results showed that mixed convection in the riser duct is an important factor for accurately extrapolating the thermo-fluid behavior in the prototype from the test results in the scale-down facilities. The system analysis code, GAMMA+, with improved heat transfer models predicted fairly well the air-cooled RCCS test data from two facilities. GAMMA+ analysis results showed that the predicted radiation fractions on the heated plate in the scale-down test conditions were larger than those on the reactor vessel in the prototype. The scaling law for air-cooled RCCS was improved by considering the mixed convection in the riser duct and the same radiation fraction on the heated plate. The mixed convection effect was calculated by the height ratio. The same radiation fraction provided a conservative extrapolation from the scale-down test results.

The analysis of air-cooled heat exchanger for Reactor Cavity Cooling System (RCCS) on Experimental Power Reactor (RDE) design

The National Nuclear Energy Agency (BATAN) is designing the Experimental Power Reactor (RDE). The Reactor Cavity Cooling System (RCCS) is a new basic design in the nuclear power plant. It will be incorporated into High-Temperature Reactors (HTR), which is one proposed generation of IV reactors. While HTR has been designed and operated, the RCCS design is unique to the recent generation of gas-cooled reactors as a passive decay heat removal system. However, many aspects of the new air- or water-cooled design still need to be investigated on RCCS. The temperature, mass flow, configuration of the riser, and number of the risers on the water-cooled heat exchanger system which is used is a necessary test. The RCCS designs have not been studied for the RDE design. The heat-transfer and fluid-flow on the RCCS facility have not been investigated for the air-cooled cavity cooling system. The aim of this study is to examine the condition of the air-cooled heat exchanger of RCCS for RDE 10 MWth design. The pressure, temperature, and velocity with certain diameters of cooling pipes are computed with FLUENT 6.3 computational fluid dynamics (CFD) software package under a mass flow rate variation. The pressure, the flow velocity and delta temperature outlet were examined on the RCCS process. The air-cooled heat exchanger in RDE basic design was found to attain safe operation.

Comparative methodology between actual RCCS and downscaled heat-removal test facility

Previously, a reactor cavity cooling system (RCCS) has been reported with passive safety features comprising of two continuous closed regions, namely an ex-reactor pressure vessel region and cooling region with a heat-transfer surface to ambient air. The novel shape of the RCCS allows it to efficiently remove heat released from the reactor pressure vessel (RPV) via thermal radiation and natural convection. The RCCS design significantly reduces the possibility of losing the heat sink for decay heat-removal during nuclear accidents including a station blackout by employing air as a working fluid and ambient air as ultimate heat sink. RCCS has the potential to stably and passively remove heat released from the RPV and decay heat following a reactor shutdown. The RCCS achieved a heat-removal rate of approximately 3 kW/m2. On the contrary, the heat fluxes from the RPV surface of the high temperature engineering test reactor and commercial high temperature gas-cooled reactors are 1.23–2.46 kW/m2 and approximately 3.0 kW/m2, respectively. In the previous report, the authors changed the adiabatic boundary conditions and considered the heat dissipation effect from the RPV region to ground through the RCCS wall via heat conduction; therefore, the authors could improve the system’s heat-removal capability to increase its thermal reactor power level. Moreover, considering the possibilities for doubling the heat-transfer areas and increasing the emissivities, heat flux removed by the RCCS could potentially reach 7.0 kW/m2. Herein, the authors conduct a comparative methodology between an actual RCCS and a downscaled heat-removal test facility.

Assessment of RELAP5 code model to simulate U-shaped heat pipe performance for heat sink

One of important component in a high temperature gas cooled reactor (HTGR) is reactor cavity cooling system (RCCS). The function of the RCCS is to remove the heat going out from the reactor pressure vessel. On the other hand, the use of passive system is highly considered in the nuclear reactor design, including in the HTGR design. The heat pipe is one of the passive apparatus. This current study deals with the research on the use of heat pipe to the RCCS design. As part of the study, the performance of U-shaped heat pipe will be assessed both through an experiment and numerical simulation methodology. The objective of this study is to assess the validity of the RELAP5 code model to simulate a U-shaped heat pipe performance. A small U-shaped heat pipe with fins in the condenser part is modelled using the RELAP5/SCDAP code. The result of simulation shows that the heat pipe could removed the heat received in the evaporator to the environment through the condenser of that heat pipe. The phenomena of evaporation and condensation are identified during heat removal. However, the steady state condition is not obtained until the end of the calculation. The calculated thermal resistance of the heat pipe is approx. 0.01258 °C/W. It is concluded that the model could simulate reasonably the heat pipe performance. The model needs to be improved, especially for the finned tube condenser.

Improvement of heat-removal capability using heat conduction on a novel reactor cavity cooling system (RCCS) design with passive safety features through radiation and natural convection

A previously-reported concept of reactor cavity cooling system (RCCS) with passive safety features consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to efficiently remove the heat released from the reactor pressure vessel (RPV) through thermal radiation and natural convection. Employing air as a working fluid and ambient air as an ultimate heat sink, the novel RCCS design strongly reduces the possibility of losing the heat sink for decay-heat-removal during nuclear accidents including a station blackout. The RCCS could stably and passively remove the heat released from the RPV and the decay heat after reactor shutdown. The previously-reported heat-removal rate of the RCCS was approximately 3 (kW/m2). The heat flux from the RPV surface of the High Temperature engineering Test Reactor (HTTR) is almost in the same range; 1.23–2.46 (kW/m2). In this paper, the authors address an improvement of heat-removal capability by considering potential of heat leakage due to heat conduction through the RCCS wall aimed at increasing a thermal reactor power level. Under the assumption of doubling the RCCS wall heat transfer area, a heat-flux removed by the RCCS could be doubled, such as approximately 6.2 (kW/m2).

New reactor cavity cooling system (RCCS) with passive safety features: A comparative methodology between a real RCCS and a scaled-down heat-removal test facility

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features and without electrical or mechanical driving requirement is proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region (RPV region) and a cooling region with a heat-transfer surface to ambient air assumed to be 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV through radiation and natural convection. Employing air as the working fluid and ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay-heat-removal. In the paper, we address a comparative methodology between a real RCCS (REAL) and a scaled-down heat-removal test facility (SCALE). As an example, the heat flux by radiation of (SCALE) can be the same as that of (REAL) because the two have the same temperatures of the RPV and the RCCS, TRPV(REAL)=TRPV(SCALE), TRCCS(REAL)=TRCCS(SCALE), and the same view and configuration factors. Next, we also conduct comparisons on natural convection using the Grashof number, GrrRCCS-rRPV. Here, the ratio of (SCALE) to (REAL) in terms of characteristic length is defined as (rRCCS-rRPV)(SCALE)(rRCCS-rRPV)(REAL)=1x. When the working fluid in (SCALE) can be pressurized up to x32 times that of (REAL), (SCALE)’s Grashof number, GrrRCCS-rRPV, can be the same as that of (REAL). Moreover, (SCALE)’s heat flux on the RPV surface as the experimental conditions for obtaining the same Grashof number as that of (REAL) can be determined. The difference between (SCALE) and (REAL) in terms of the heat flux on the RPV surface is (x−1) times (REAL)’s heat flux by natural convection. Thus, (SCALE) will obtain valuable experimental data for demonstrating the new RCCS’s performance.

New reactor cavity cooling system having passive safety features using novel shape for HTGRs and VHTRs

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed for high temperature gas cooled reactors (HTGRs) and very high temperature reactors (VHTRs). The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40(°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting due to overheating the fuels. The simulation results from a commercial CFD code, STAR-CCM+, show that the temperature distribution of the RCCS is within the temperature limits of the structures, such as the maximum operating temperature of the RPV, 713.15(K)=440(°C), and the heat released from the RPV could be removed safely, even during a loss of coolant accident (LOCA). When the RCCS can remove 600(kW) of the rated nominal state even during LOCA, the safety review for building the HTTR could confirm that the temperature distribution of the HTTR is within the temperature limits of the structures to secure structures and fuels after the shutdown because the large heat capacity of the graphite core can absorb heat from the fuel in a short period. Therefore, the capacity of the new RCCS design would be sufficient for decay heat removal.

Assessment of MARS-GCR Code for a New Reactor Cavity Cooling System Design

Seoul National University (SNU) proposed a new concept of a reactor cavity cooling system (RCCS), which is a critical safety feature in high-temperature gas-cooled reactors. To provide reasonable experimental data for the code assessment and evaluate the feasibility of the proposed system, performance and integrity were tested by separate-effects test apparatuses and a reduced-scale mockup facility named RCCS-SNU. Calculations were performed using the MARS-GCR code for the validation of its capability to simulate multidimensional behavior, natural convective heat transfer, radiative heat transfer, etc. This assessment showed that the MARS-GCR code reasonably predicts the characteristics of the radiative heat transfer in the cavity and the forced convective heat transfer through the air-cooling pipes. However, the study showed deviation in the simulation of heat transfers that occur inside the cavity and water pool, especially the thermal stratification phenomenon. As a result, it was concluded that applying the system code with coarse node, MARS-GCR had certain limitations in the simulation of local phenomena.

Evaluation and optimization of General Atomics’ GT-MHR reactor cavity cooling system using an axiomatic design approach

The development of the Generation IV (Gen-IV) nuclear reactors has presented social, technical, and economical challenges to nuclear engineering design and research. To develop a robust, reliable nuclear reactor system with minimal environmental impact and cost, modularity has been gradually accepted as a key concept in designing high-quality nuclear reactor systems. While the establishment and reliability of a nuclear power plant is largely facilitated by the installment of standardized base units, the realization of modularity at the sub-system/sub-unit level in a base unit is still highly heuristic, and lacks consistent, quantifiable measures. In this work, an axiomatic design approach is developed to evaluate and optimize the reactor cavity cooling system (RCCS) of General Atomics’ Gas Turbine-Modular Helium Reactor (GT-MHR) nuclear reactor, for the purpose of constructing a quantitative tool that is applicable to Gen-IV systems. According to Suh's axiomatic design theory, modularity is consistently represented by functional independence through the design process. Both qualitative and quantitative measures are developed here to evaluate the modularity of the current RCCS design. Optimization techniques are also used to improve the modularity at both conceptual and parametric level. The preliminary results of this study have demonstrated that the axiomatic design approach has great potential in enhancing modular design, and generating more robust, safer, and less expensive nuclear reactor sub-units.