Reactor Cavity Cooling System Research Articles

Numerical study of the flow and heat transfer characteristics in a scale model of the vessel cooling system for the HTTR

The reactor cavity cooling system (RCCS) is a passive reactor safety system commonly present in the designs of High-Temperature Gas-cooled Reactors (HTGR) that removes heat from the reactor pressure vessel by means of natural convection and radiation. It is one of the factors responsible for ensuring that the reactor does not melt down under any plausible accident scenario. For the simulation of accident scenarios, which are transient phenomena unfolding over a span of up to several days, intermediate fidelity methods and system codes must be employed to limit the models’ execution time. These models can quantify radiation heat transfer well, but heat transfer caused by natural convection must be quantified with the use of correlations for the heat transfer coefficient. It is difficult to obtain reliable correlations for HTGR RCCS heat transfer coefficients experimentally due to such a system’s size. They could, however, be obtained from high-fidelity steady-state simulations of RCCSs. The Rayleigh number in RCCSs is too high for using a Direct Numerical Simulation (DNS) technique; thus, a Reynolds-Averaged Navier–Stokes (RANS) approach must be employed. There are many RANS models, each performing best under different geometry and fluid flow conditions. To find the most suitable one for simulating an RCCS, the RANS models need to be validated. This work benchmarks various RANS models against three experiments performed on the HTTR RCCS Mockup by the Japanese Atomic Energy Agency (JAEA) in 1993. This facility is a 1/6 scale model of a vessel cooling system (VCS) for the High Temperature Engineering Test Reactor (HTTR), which is operated by JAEA. Multiple RANS models were evaluated on a simplified 2d-axisymmetric geometry. They were found to reproduce the experimental temperature profiles with errors of up to 22% for the lowest temperature benchmark and 15% for the higher temperature benchmarks. The results highlight that the pragmatic turbulence models need to be validated for high Rayleigh natural convection-driven flows and improved accordingly, more publicly available experimental data of RCCS resembling experiments is needed and indicate that a 2d-axisymmetric geometry approximation is likely insufficient to capture all the relevant phenomena in RCCS simulations.

Effect of inlet throttling on thermohydraulic instability in a large scale water-based RCCS: An experimental study

The objective of the present experimental study is to investigate the effect of inlet throttling on the thermohydraulic stability of a large scale water-based Reactor Cavity Cooling System (RCCS). The test was performed using the water-based Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne, which represented a ½ axial scale and 12.5° sector slice of the full scale Framatome 625 MWt SC-HTGR RCCS concept. A two-phase steady state was first established through direct condensate refill, followed by increased inlet throttling over 10 stages, corresponding to a loss coefficient K over the range of 0.05–653. With the inlet throttling gradually increased, the system experienced a unique transition process between stabilization and destabilization. Through a stability analysis, three instability mechanisms were identified in the present test, including a compound mechanism due to both natural circulation oscillations (NCOs) and density wave oscillations (DWOs), Type-II DWOs, and geysering.

On the role of low frequency modes in the thermal and momentum mixing in parallel plane jets

Parallel plane jet mixing has been the subject of extensive study and its physics are characterized well in literature. However, the physics of parallel plane jet mixing in relatively small confined regions is not well known. In this study we investigate the effects of re-circulatory flow on thermal and momentum mixing in confined parallel plane jets. Studying confined jet interaction is important for a variety of engineering applications. One of these applications is sodium cooled fast reactors (SFR). It is common in SFRs to have coolant stream mixing in a confined domain, often the upper plenum. The Reactor Cavity Cooling System (RCCS) facility at the University of Michigan models this problem in a simplified enclosure with isothermal parallel plane jets as the incoming coolant streams. A Large Eddy Simulation (LES) of this simplified plenum was developed using optimal boundary conditions, proven mesh convergence, and consistent resolution. This model was then validated against experimental results from the RCCS facility. After model validation, flow structures in the jet interaction were discerned by means of Proper Orthogonal Decomposition (POD). Comparison of 3D POD to a clipped 2D POD region reveals that re-circulatory flow can lead to dominant oscillations within the jet mixing region. Subsequently, Power Spectrum Density (PSD) and Continuous Wavelet Transforms (CWT) were used to ascertain frequency-based phenomena that might occur in this mixing region. Frequency analysis uncovered a dominant low-frequency oscillation in both velocity and temperature PSDs. These frequencies fall within the known ranges of thermal striping in sodium. The data generated in this study can be used for the benchmarking of less computationally intensive approaches. This data will be made available upon request to ebm5351@psu.edu.

Assessing the Structure-Based Turbulence Model Performance for Thermal Striping Applications Using Symmetric Jet Experiments

Turbulent mixing of coolant streams can result in an oscillatory mixing phenomenon called thermal striping. These fluctuations have the potential to lead to anticipated thermal fatigue failures in advanced nuclear reactors. To predict thermal striping, robust and computationally affordable modeling tools that are capable of accurately representing complex turbulence are needed. Hybrid turbulence approaches, such as detached-eddy simulation and scale-adaptive simulation, have shown some success in resolving complex unsteady turbulence for massively separated flows, however the applicability of these models to internal flows is limited. A STRUCTure-based (STRUCT) second-generation Unsteady Reynolds-Averaged Navier–Stokes turbulence model was recently proposed at the Massachusetts Institute of Technology to robustly extend the applicability of hybrid closures. In this work, the STRUCT model is evaluated using experimental data taken at the Reactor Cavity Cooling System separate-effects test facility at the University of Michigan. The experiments observed the interaction of parallel symmetric rectangular jets, and include measurements for mean profiles of velocity and Reynolds stresses. In the present work, the simulation results are assessed against mean profiles of velocity and Reynolds stresses, demonstrating the ability to reproduce the unsteadiness of the jets in close agreement with the measurements at considerably reduced computational cost.

Thermal-hydraulic behavior simulations of the reactor cavity cooling system (RCCS) experimental facility using Flownex

The scaled water-cooled Reactor Cavity Cooling System (RCCS) experimental facility reproduces a passive safety feature to be implemented in Generation IV nuclear reactors. It keeps the reactor cavity and other internal structures in operational conditions by removing heat leakage from the reactor pressure vessel. The present work uses Flownex one-dimensional thermal-fluid code to model the facility and predict the experimental thermal-hydraulic behavior. Two representative steady-state cases defined by the bulk volumetric flow rate are simulated (Re = 2,409 and Re = 11,524). Results of the cavity outlet temperature, risers' temperature profile, and volumetric flow split in the cooling panel are also compared with the experimental data and RELAP system code simulations. The comparisons are in reasonable agreement with the previous studies, demonstrating the ability of Flownex to simulate the RCCS behavior. It is found that the low Re case of 2,409, temperature and flow split are evenly distributed across the risers. On the contrary, there's an asymmetry trend in both temperature and flow split distributions for the high Re case of 11,524.

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.

MELCOR integrated severe accident code application to safety assessment of high-temperature gas-cooled reactors

MELCOR is an integrated thermal hydraulics, accident progression and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (U.S. NRC) since the early 1980 s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts over the past decades have expanded its application space to non-LWR reactor concepts. Current MELCOR development efforts have been focused on providing the U.S. NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC’s near-term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. In this paper, we describe the generic plant model developed to demonstrate MELCOR capabilities to perform high-temperature gas reactor (HTGR) safety evaluations. This model represents a TRi-structural ISOtropic particle fuel (TRISO) pebble bed HTGR with a primary system rejecting heat to a recuperative heat exchanger. Surrounding the reactor vessel is a reactor cavity contained within a confinement volume cooled by the Reactor Cavity Cooling System (RCCS). A range of demonstration calculations are performed to evaluate the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, primary system heat removal and reactor vessel decay heat removal, and confinement cooling.

Investigation into different 1D/3D co-simulation methodologies applied to a natural circulation loop

1D/3D co-simulation is a promising approach for the modelling and simulation of complex systems. These systems are typically computationally intensive. By combining the desired features of a one-dimensional (1D) systems code and a three-dimensional (3D) Computational Fluid Dynamics (CFD) code, the required resources are optimized whilst still retaining an acceptable level of accuracy. Co-simulation has been applied in various domains and scenarios, often without a comprehensive study on the impact to simulation- stability, -time, and results. The purpose of this study is to explore the behaviour of some of the strategies by applying them to the simulation of the natural circulation occurring in a simplified Reactor Cavity Cooling System (RCCS). Four different co-simulation approaches were explored to investigate the radiative- and convective- heat transfer as well as the natural circulation phenomena occurring in a RCCS facility. Typically, in these four modelling approaches a 3D methodology was employed to model the radiation and convection heat transfer in the cavity between the heater wall and the outer wall of the riser. The 1D methodology was employed to model the remainder of the RCCS system. The 1D systems code, Flownex, and the 3D CFD code Ansys Fluent were used to perform the 1D/3D co-simulations. Each of the codes modelled a relevant part of the RCCS facility and the models were coupled by explicitly exchanging boundary conditions or relevant data between the 1D and 3D regions. The validity of the coupled 1D/3D approaches was investigated by comparing them with a full 1D simulation.

Turbulence model assessment and heat transfer phenomena inside a rectangular channel under forced and mixed convection

A very high-temperature gas-cooled reactor (VHTR) is an attractive option for hydrogen production due to its high operating temperature. The reactor cavity cooling system (RCCS), a passive safety system of the VHTR, surrounds the reactor vessel with rectangular riser channels and removes decay heat from it. Under emergency operation conditions such as a decrease in the flow rate of the RCCS, the flow condition inside the RCCS riser can become a mixed convection condition accompanying heat transfer deterioration. To design and operate the VHTR with reliable safety, accurate predictions for various convective heat transfer conditions in RCCS riser channels are required. In this study, by comparing the results of an air flow visualization experiment, the predictability of the turbulence models was assessed for the convective heat transfer and complex flow characteristics near the corner region of the rectangular duct confirmed from the experimental study. In this paper, the heat removal rate through the test section of the visualization experiment was quantified under various convective heat transfer conditions, and local temperature measurement data from the experiment were presented to describe the heat transfer deterioration phenomena inside a rectangular riser in turbulent mixed convection. Linear and nonlinear eddy viscosity models were evaluated by comparing the heat transfer coefficient, local temperature, local flow structure including secondary flow, and turbulence quantities obtained from the experiment. Furthermore, improvements to turbulence models for enhancing the prediction of mixed convection heat transfer in rectangular riser ducts were proposed.

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.

Study on the DLOFC accident of the GEMINI+ conceptual design of HTGR reactor with MELCOR and SPECTRA

The work presented in this paper was performed within the Euratom Horizon 2020 GEMINI Plus project. Behavior of the HTGR reactor under severe accident conditions was investigated and the maximum fuel temperature was observed. Due to application of the TRISO particles and SiC layers in the fuel element, no damage of the fuel is expected up to 1600°C. Under the cooperation in the project between Nuclear Research Group (NRG) and National Centre for Nuclear Research (NCBJ) a code-to-code calculations were carried out between the SPECTRA and MELCOR codes. SPECTRA code, developed by the NRG is a thermal hydraulic analysis code and MELCOR 2.1.6342 used by NCBJ developed by SANDIA National Laboratory is fast running severe accident code. Both codes have already HTGR specific models build in. The following accident was analyzed and will be presented: Depressurized Loss of Forced Circulation (DLOFC) with 65 mm break at the top of reactor vessel. The scenario was calculated applying following sets of assumptions: best estimate and conservative. Plant behavior was analyzed including primary and secondary side of the reactor. As the results of applying conservative assumptions, it was found that fuel temperature excides the acceptable limit of 1620°C. Therefore, changes in the core design were proposed by project participants. Analyses of the new core showed acceptable temperatures. In the paper the results of code-to-code comparison are presented. Both codes have shown a good agreement of presented following characteristics on maximum fuel temperature, relative power and Reactor Cavity Cooling System power, primary pressure and break flow.

Improved natural convection heat transfer correlations for reactor cavity cooling systems of high-temperature gas-cooled reactors: From computational fluid dynamics to Pronghorn

The reactor cavity cooling system (RCCS) is a common reactor safety system in high-temperature gas-cooled Reactors (HTGR) that removes heat from the reactor pressure vessel (RPV) by radiation (∼80%) and natural convection (∼20%). For the simulation of accident scenarios of HTGRs, intermediate fidelity and system codes models must be employed to limit the models’ execution time. While an accurate quantification of the radiative heat transfer is available in these models, the quantification of natural convection must rely on correlations of questionable accuracy for the Nusselt number. Commonly used correlations are based in experiments performed at low Rayleigh numbers and/or using isothermal walls in simplified geometries. This work improves on the accuracy of natural convection heat transfer correlations to support HTGR designs.These correlations include both local and average Nusselt numbers as a function of the global Rayleigh number, the local Rayleigh number, and the temperature profile at the hot wall of the RCCS. In the absence of dedicated experiments and the difficulty of performing high-fidelity simulations at realistic Rayleigh numbers, the data to fit the correlations are generated with computational fluid dynamics (CFD) using Reynolds Averaged Navier-Stokes (RANS) models. First, a careful selection of the RANS turbulence model is performed by comparing the results obtained with different RANS turbulence models against high-fidelity simulations of natural convection at Ra 1×1011 in a rectangular cavity. Next, the selected model is used to perform simulations of an HTGR cavity at different high Rayleigh numbers ∈[6.1×1011,2.9×1013] to encompass several HTGR designs, assuming an isothermal RPV wall. The results obtained are used to fit a correlation for the average and space-varying Nusselt number as a function of the global and local Rayleigh numbers via a sparsity-promoting, least-squares method. The selected RANS model is then used to perform simulations of a PBMR-400 (Pebble Bed Modular Reactor) HTGR cavity with the temperature profiles at the RPV wall obtained during a PLOFC (Pressurized loss of forced cooling) transient. We use the results obtained to fit a temperature-dependent correction to the space-varying Nusselt number with the sparsity-promoting, least-squares method. The results obtained in this work enable system-level codes, such as Pronghorn, to perform higher-fidelity simulations of the heat exchange process in the RCCS while still maintaining a low computational cost.

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.

Review of research using analogy concept for thermal hydraulic and severe accident experiments

This paper introduces the experimental method of using mass transfer systems for heat transfer studies based on the analogy between heat and mass transfer. The copper electroplating system, composed of copper electrodes submerged in an aqueous solution of H2SO4 and CuSO4, is employed as the mass transfer system. The cupric ions transferred from anode to cathode simulate heat transfer. Applications of the experimental method are presented regarding the natural convection of the oxide pool in the IVR-ERVC (In-Vessel Retention and External Reactor Vessel Cooling) and the highly buoyant mixed convective flow conditions of the RCCS (Reactor Cavity Cooling System). The experimental method is extended to two phase flows using the reduced hydrogen at the cathode simulating the vapor bubbles. Some experiments simulating the CHF (Critical Heat Flux) problems are introduced. The paper aims at introducing the analogy experimental methodology as the preliminary quantitative experimental tools.

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

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.

The RCCS Thermal Analysis during the Station Blackout Accident of RDE

RDE is designed based on HTGR technologies; it has an active, passive and inherent safety system. Defence in depth philosophy is conducted. The heat transferred from the reactor vessel to the structure of the containment is equipment in the RDE has a name is a reactor cavity cooling system (RCCS). During normal operation and protecting the reactor containment structure in the event of failure of all active cooling systems, the RCCS designed to take fulfill this role by preserving the vessel reactor under the maximum allowable temperature. Concerning heat dissipation design, RCCS assessment and resolution are considered as major factors in determining the maximum power level of heat dissipation. Active and passive mode is the equipment on RDE RCCS. This paper discusses the calculation of RCCS thermal analysis during the station blackout (SBO) accident. SBO occurred and the heat removal is conducted by RCCS when all of the active cooling systems were failed. Conditions will be safe if the RPV temperature must be maintained below 65°C. There is no electricity from diesel generator supplied to the primer compressor, it will the SBO accident. The methodology used is based on the calculation mathematical model of the pebble bed modular reactor heat transfer capability of the RCCS in the passive mode. The passive mode works during the failure of active mode, and the heat is released through the cavity by natural convection. The RCCS is capable to withdraw the heat at 50.54 kW per hour and keep the reactor pressure vessel temperature below 65°C at normal operation and below 125°C during the station blackout.

Heat transfer simulation in reactor cavity cooling system for experimental power reactor (RDE)

To enhance the safety of experimental power reactor (RDE), Reactor Cavity Cooling System (RCCS) is required to support primary helium coolant. RCCS can be utilized for both normal condition or under accident. During the reactor accident with no power supplies and cooling pump down, RCCS works in a natural-circulation by using air in cavity and water in the cooling pipe. In this study, the process of radiation heat transfer from Reactor Pressure Vessel (RPV) to Water Cooled Pipe (WCP) was simulated. The natural circulation of air cavity and around the WCP was studied in three dimensional numerical model. The analyses are performed with Fluent code, which is a computational fluid dynamic code. Three parameter conditions of simulation were studied to compare the difference of them, so in this primary study, it can be known the influence of each parameter to the obtained results and it can be considered for development of future safety simulation in RDE reactor. From the simulation result, the temperature at bottom air cavity is higher than the top of the cavity. The temperature profile of water in the pipe also changes because of influence the heat transfer from Reactor Pressure Vessel. Gravity and radiation are very important parameter inputs to simulations in order to approach the real conditions.

Local flow structure and turbulence quantities inside a heated rectangular riser in turbulent forced and mixed convection heat transfers

Heat transfer deterioration in turbulent mixed convection affects the performance and safety of various pieces of industrial equipment. The reactor cavity cooling system (RCCS), a passive safety system in a very high temperature gas-cooled nuclear reactor (VHTR), incorporates rectangular riser channels to remove residual heat emitted from the reactor vessel, where turbulent mixed convection may occur under some emergency operation conditions. Understanding the heat transfer phenomena in the RCCS riser duct is, therefore, important to ensure the safety of the VHTR. To investigate the heat transfer phenomena inside a heated rectangular duct, a visualization experiment was carried out to obtain the local flow structure and turbulence quantities in turbulent forced and mixed convection heat transfers. Using the particle image velocimetry method, the velocity fields inside the heated test section were captured and time-averaged velocity distributions and root-mean-square velocity fluctuation distributions were obtained under various experimental conditions. Inside a rectangular test section, the flow had a nonaxisymmetric Reynolds shear stress distribution, related to secondary flows near the corner of the test section. According to the overall distributions of the local velocity and turbulence quantities under various experimental conditions, the mechanism of the heat transfer deterioration inside a heated rectangular riser was investigated in relation to the distribution of the Reynolds shear stress. Inside the rectangular test section with the intense buoyancy force, flow laminarization near the corner was preceded by that of the center with decreases in secondary flows and Reynolds shear stress distribution near the corner, which indicate a transition from forced to mixed convection. According to these results of the visualization experiments, it was estimated that the heat transfer deterioration in the rectangular riser would be affected by the aspect ratio of the test section.

Whole-core transient analysis methodology for prismatic high temperature gas reactors

A new transient thermalhydraulic method for simulating prismatic high temperature gas cooled reactors (HTGRs) during a loss-of-forced-circulation (LOFC) accident is presented. Whole-core, three-dimensional temperature profiles are computed at an individual fuel pin and lattice assembly block levels. A method that computes the fluid mass, velocity (momentum), and energy throughout a transient is implemented using a well-documented, semi-implicit pressure-correction scheme. Models for reactor cavity cooling system (RCCS) heat transfer and decay heat generation are also included. Pressurized (P-LOFC) and de-pressurized (D-LOFC) accidents are simulated. This method provides two significant advantages over available thermalhydraulic analysis techniques. The first is its ability to compute whole-core results and capture the important transient core-level phenomena such as bypass flow and heat redistribution into the reflector assemblies after reactor SCRAM. The second is its ability to compute each fuel assembly in detail, computing the heat rates and temperature profiles for every fuel pin, graphite, and coolant channel. These features produce realistic, 3-D transient results for prismatic HTGRs during a LOFC. A RELAP model is also developed and implemented for the HTGR. The limitations of existing methods for the simulation of the prismatic fuel are identified and the need for and novelty of the present method are highlighted.

CFD Investigation of MHTGR Natural Circulation and Decay Heat Removal in P-LOFC Accident

One of the key postulated accidents in a high-temperature gas-cooled reactor (HTGR) is the pressurized loss of forced circulation (P-LOFC) of the primary loop, which can be triggered by its primary helium circulator trip or turbine trip. If the reactor shutdown cooling system (SCS) fails during a P-LOFC accident, part of the reactor decay heat is absorbed by the reactor core materials and the rest removed by the reactor cavity cooling system (RCCS). In the extended period of P-LOFC accident compound with SCS failure, the core decay heat is supposed to be removed by conduction, natural circulation, convection, and radiation. A three-dimensional (3-D) computational fluid dynamics (CFD) simulation was performed in this research to study the long-term heat removal mechanisms in the General Atomics’ Modular High Temperature Gas-cooled Reactor (MHTGR) design in a P-LOFC accident. The reactor core temperature distribution and flow field were obtained at different decay power levels. The sensitivity of the natural circulation flow to the bypass gap width was investigated. The natural circulation flow intensity is relatively weak but very sensitive to the width of the bypass gaps.