Heater Rod Research Articles

An experimental study of the CHF in a square channel with non-uniformly heated rod under low pressure and low flow conditions

Experimental research on critical heat flux (CHF) was carried out in a vertical square channel with non-uniformly heated rod at low pressure and low flow conditions. The influence of inlet subcooling on CHF was also investigated. The results showed that the CHF position of the non-uniformly heat flux distribution mainly appeared in the middle and lower part of the heater rod. More significantly, it was found that CHF was enhanced with the rise of the inlet subcooling. The CHF data measured in this experiment were also assessed with the commonly used CHF empirical correlations suitable for low pressure and low flow. The results indicated that the general correlations predicting for CHF of uniform heat flux distribution were not applicable for prediction CHF of non-uniform heat flux distribution and these correlations usually overestimated the experimental values. In addition, The F-factor approach proposed by Tong was used to predict the experimental data and the error predicted by F-factor were mostly between 30% and 50%. Finally, F-factor was modified based on experimental data and the error between the predicted value and experimental results was between 0 and 30 %, and most of the data was between 0 and 20 %.

RELAP5-3D validation studies based on the High Temperature Test facility

In the spring and summer of 2019, experiments were conducted at the High Temperature Test Facility (HTTF) that form the basis of an upcoming high-temperature gas-cooled reactor (HTGR) thermal hydraulics (T/H) benchmark. HTTF is an integral effects test facility for HTGR T/H modeling validation. This paper presents RELAP5-3D models of two of those experiments: PG-27, a pressurized conduction cooldown (PCC); and PG-29, a depressurized conduction cooldown (DCC). These models used the RELAP5-3D model of HTTF originally developed by Paul Bayless as a starting point. The sensitivity analysis and uncertainty quantification code, RAVEN was used to perform calibration studies for the steady-state portion of PG-27. We developed four PG-27 calibrations based on steady-state conditions. These calibrations all used an effective thermal conductivity equal to 36 % of the measured thermal conductivity, but they differed with respect to the frictional pressure drops and radial conduction models. These models all captured the trends in steady-state temperature distributions and transient temperature behavior well. All four calibrations show room for improvement in predicting the transient temperature rise. The smallest error in temperature rise during the transient was a 21 % underprediction, and the largest was a 48 % underprediction. The errors in transient temperature rise are largely a result of a mismatch in power density between the RELAP5-3D model and the experiment due to the location of active heater rods along the boundary between heat structures in the model. The best of these calibrations was applied to PG-29 to model the DCC. Once again, temperatures during the transient were underpredicted but trends in temperature were captured. The RELAP5-3D model captured trends in the data but could not reproduce measured temperatures exactly. This result is not attributed to deficiencies in the experimental data or to RELAP5–3D itself. Rather, this result likely arises due to the some of the assumptions and decisions made when the RELAP5-3D model was first developed, prior to the execution of HTTF experiments. An agreement in prediction of temperature trends but challenges reproducing HTTF temperatures within measurement uncertainty is consistent with previous analyses of HTTF in the literature. Future RELAP5-3D validation activities centered around HTTF may be able to provide greater insight into the code’s capabilities for HTGR modeling with a more finely nodalized model.

Simulation of natural convection flow for vertical heated rod by using ANSYS/Fluent

The decay heat removal by natural convection is very important in case of Station blackout (SBO) of nuclear reactor. The computational fluid dynamic (CFD) is helpful to simulate the flow and temperature field. However, the CFD simulation models need to be validated by the experimental data. Thus, in this report, the Anslys/Fluent is applied to simulate the natural convection induced by single heater rod. The vertical heated rod with a diameter of 12 mm and a length of 225 mm is immersed at the center of a vertical pipe made of acrylic with a diameter of 146 mm and a height of 500 mm. In this simulation, the coupled scheme algorithm is also applied. Regarding to experimental method, the optical method such as Particle Image Velocimetry (PIV) was applied for 2-dimensional velocity distribution and pointwise temperature measurement. As a result, the predicted flow and temperature field are well agreed with experimental data. Accordingly, the thermal plumes are well estimated by ANSYS/Fluent, in which the buoyant plumes induced by different temperatures vertically go up along the heater rod's upper part until the container's isolated upper wall. The complicated flow occurs in the middle part of the container by mixing the downward flow on the top and the upward flow from the heater rod.

Experimental investigation of quench front propagation during reflooding in 5 × 5 rod bundle channel with axial uniform power distribution

This experimental study delves into the impact of inlet injection velocity, inlet subcooling, and linear power density on the quench front propagation during reflooding in 5x5 rod bundle channel with axial uniform power distribution. The investigation utilized a film boiling test facility constructed by NUclear Safety and Operation Laboratory (NUSOL) of Xi'an Jiaotong University. The findings demonstrate a discernible sensitivity of the quench front to variations in these parameters, wherein higher linear power density, reduced flow rate, and decreased coolant subcooling lead to delayed reflooding. Moreover, the study derived the surface parameters of the heater rod use self-developed Reflooding Experimental Data Analysis code (REDA). The comprehensive experimental data and empirical models generated in this study hold significant potential for enhancing the safety analysis of nuclear reactors.

A multipass catalytic reactor insert for continuous hydrogen generation from methylcyclohexane

Herein we describe a multipass Catalytic Static Mixer (CSM) for implementation in on-demand hydrogen generation. The insert is a 3D printed metal scaffold with a catalyst coating containing a multitude of channels and a hollow core for a heater rod. This next generation of CSMs offers higher throughput and increased residence times for applications involving liquid-gas or gas only flow. We have implemented this CSM as an insert in a continuous flow tubular reactor for the dehydrogenation of Liquid Organic Hydrogen Carriers (LOHCs). The catalyst coating is comprised of 5 wt.% Pt/Al2O3 and is used for the dehydrogenation of methylcyclohexane (MCH) to generate hydrogen gas as the model system. This experimental study investigates the thermal profiles of the reactor and validates the potential of the CSM technology in hydrogen generation from LOHCs.

Time in DNB experimental study on Cr coated zircaloy cladding

Since the Fukushima-Daiichi accident effort has been put in by the nuclear community to develop Accident Tolerant Fuel (ATF) cladding materials to reduce the risk of clad oxidation and, consequently, hydrogen generation and potential major accidents. Among designed materials, the use of a chromium coating on top of zirconium-alloy cladding showed reasonable results in preliminary tests, being a favorable cladding solution for Pressurized Water Reactors. This paper provides highly detailed temperature data obtained during time in departure from nucleate boiling (DNB) experiments under prototypical reactor conditions of temperature, pressure, mass flux, and heat flux profile on a Cr-coated zirconium-alloy cladded heater rod simulating the fuel pin. Data were obtained at mass fluxes from 1300 to 1700 kg/m2s, pressure of 16 MPa, and inlet temperature of 333 °C (subcooling degree of 14.4 °C). The change in dry time was obtained by increasing the spike in wall temperature necessary to shut down the power. These tests showed that increasing the spike in temperature to characterize the critical heat flux from 50 °C to 400 °C resulted in variations of the departure from nucleate boiling measured value within the uncertainty range of their measurements, hence, validating the use of a spike of 50 °C, a commonly used value in this kind of study, as an indicator of the critical heat flux. Within the dry region, the temperature was found to increase along the direction of the flow for a very short length, reaching a peak, from which it reduces linearly until the surface reaches a temperature that is no longer able to sustain a dry patch. It was noticed that the dry region can grow against the flow direction due to the cosine profile of the heat flux, and that small dryout events can occur without leading to departure from nucleate boiling, i.e., being fully suppressed. However, two consecutive events may sum up and cause the critical heat flux. Rewet time was found to increase with peak cladding temperature and decrease with mass flux. A maximum peak cladding temperature of 844 °C was found, with the surface remaining dry for approximately 14.1 s. The minimum film boiling temperature was measured during the rewet process, showing absolute values of around 580 °C, and difference between wall and inlet temperatures of about 240 °C. A post-inspection using microscope and X-ray images showed no damage to the cladding and heater structure, differently than what was noticed for the bare zircaloy in a previous study. This indicated that the presence of chromium coated cladding can improve the oxidation resistance of the cladding not only in short critical heat flux transients but also in longer events associated with major accidents, like a Loss of Coolant Accident.

Experimental study on flow boiling CHF in annulus channel under heaving conditions using simulant fluid R134a targeting nuclear reactor applications

Floating nuclear power plants have attracted significant attention as a prospective means to deliver clean energy, particularly in geographically remote areas where conventional power infrastructure is inaccessible. The ocean environment, however, can affect the thermal–hydraulic phenomena in them. Notably, in the two-phase system, Critical Heat Flux (CHF) can change due to the periodic variation of additional forces. While a few experimental studies have been conducted on CHF under heaving motion, expanding the range of test conditions is necessary to elucidate its mechanism at the operating pressure of a Pressurized Water Reactor (PWR). Hence, the present study performed an experiment on the flow boiling CHF under heaving motion. To simulate the heaving motion of the ocean environment, a heaving motion platform was designed and constructed. It can simulate the heaving motion with a maximum acceleration of 0.6 g and a period ranging from 3 to 6 s. The CHF test loop utilized refrigerant R134a as the working fluid, and the tests were conducted under the thermal–hydraulic conditions corresponding to PWR operation conditions applying a fluid-to-fluid scaling. The test section consists of a single heater rod in the annulus channel. CHF was measured under both static and heaving motion conditions, and a comprehensive parametric study was conducted on the CHF variation due to the oscillating acceleration field. The results demonstrated that, in most cases, CHF decreased compared to the static condition. The reduction in CHF due to heaving motion could reach up to 9% in the present test configuration, and the reduction ratio was proportional to the magnitude of the heaving acceleration. In addition, the effect of heaving motion was pronounced in two distinct thermal–hydraulic regions: when the critical quality approached zero and when it exceeded 0.6. Based on the experimental observations, the CHF reduction mechanism in each region was suggested.

Performance evaluation of a rotary dehumidifier with molecular sieve desiccant using coupled regeneration mode: Experimental investigation

The increasing demand for energy-efficient and environmentally friendly air conditioning systems has led to the development of innovative technologies such as desiccant cooling systems. This study presents a comprehensive experimental analysis of the dehumidification and thermal performance of a rotary dehumidifier with molecular sieve desiccant designed to effectively remove moisture from the air by utilizing coupled regeneration mode (complete waste heat liberated out of condenser and electric rod heat). Various experiments were conducted to investigate different performance parameters including moisture removed from air by adsorption, moisture added to air by regeneration, dehumidification effectiveness, regeneration effectiveness, moisture removal capacity (MRC), regeneration rate (RR) at different process air inlet temperatures ranging (28–34 °C) and at different regeneration air inlet temperatures ranging (55–62 °C). These experiments were conducted at same air velocity for both adsorption and regeneration segments of rotary solid desiccant wheel using molecular sieve. The results showed that the performance of the desiccant wheel was affected by inlet temperature and humidity ratio for adsorption and regeneration segments. The findings emphasize the system's ability to provide efficient dehumidification and cooling while simultaneously promoting sustainable energy practices. Moreover, the waste heat recovery system significantly contributes to the overall efficiency improvement of the rotary dehumidifier with molecular sieve desiccant, making it an environmentally friendly alternative to conventional air conditioning technologies.

Safety considerations related to boron redistribution during Loss-of-Coolant-Accident initiated beyond design basis events for the NuScale power module

The authors, as staff at the U.S. Nuclear Regulatory Commission (NRC), evaluated the potential consequences of boron dilution under beyond design basis conditions in the NuScale Power Module to inform NRC staff’s safety evaluation report for the NuScale design certification review. The authors identified the small break loss of coolant accident with coincident failure of control rods to insert as the scenario with the most adverse boron dilution and reactivity insertion consequences. The authors evaluated the sequence of events for the scenario to determine which conditions or phenomena could produce the potential for a core flow surge to yield core damage because of a super-prompt reactivity excursion. The authors screened the conditions and phenomena that could potentially cause a core flow surge based on the dynamic speed of the incursion compared to key reactivity feedback mechanisms: thermal–hydraulic coupling via rod heat transfer and convective mixing. Based on the screening process, the authors were able to identify two processes that could potentially lead to a power excursion. However, the authors’ analysis relies heavily on assumptions regarding convective mixing, making this an important phenomenon worth careful consideration in any evaluation of boron redistribution.

Sensitivity Analysis of Transient Critical Heat Flux by RIA Under High-Pressure Flow Boiling Conditions in TRTL

A reactivity-initiated accident (RIA) is a design-basis accident under which critical heat flux (CHF) is likely to be exceeded. The operational margin for RIAs is currently determined using steady-state CHF lookup tables, which provide conservative estimates relative to transient CHF phenomena. The Transient Reactor Test Loop (TRTL) facility at Oregon State University is capable of performing out-of-pile rapid heating experiments representative of a RIA at conditions representative of a pressurized water reactor (PWR). To further our understanding of and ability to predict transient CHF under PWR conditions, we performed a sensitivity analysis on a RELAP5-3D model of the TRTL facility coupled to the RAVEN code framework to define a proposed experimental test matrix to be performed at the TRTL facility. We then implemented a flow boiling CHF correlation into RELAP5-3D and performed a secondary sensitivity analysis inspecting the impact of the built-in RELAP5-3D CHF and heat transfer multipliers on both the prediction of CHF and key safety parameters, such as peak cladding temperature and heat flux. The results show that the multiplier with the highest influence toward the prediction of CHF occurrence and the safety parameters is the transient CHF multiplier. Operational performance envelopes have been developed for each of the test matrix cases and will be used for validation once the experiments are performed. The TRTL facility is currently performing shakedown testing to verify system performance prior to proceeding with the experimental campaign. Restart testing results include pump curve restart testing, pressure tests, and heater rod thermocouple transients.

Code Benchmark of the HTTF Pressurized Conduction Cooldown Test Using SAM

The High Temperature Test Facility (HTTF) at Oregon State University is an integral system test facility to simulate postulated reactor transients of prismatic high-temperature gas-cooled reactors(HTGRs). A series of test campaigns was launched, providing abundant test data that could be used to benchmark reactor system analysis codes like the System Analysis Module (SAM). In this study, a SAM model of the facility is developed based on the two-dimensional (2D) ring model approach. All components including the ceramic matrix, graphite heaters, helium coolant channels, core barrel, upcomer, pressure vessel, and reactor cavity cooling system are modeled as concentric cylindrical rings. The model is used to simulate one of the benchmark problems—Pressurized Conduction Cooldown (PCC)—within the scope of the Organisation for Economic Co-operation and Development Nuclear Energy Agency International HTTF Benchmark. The simulations consist of two parts. In the first part, operating and boundary conditions as well as thermophysical properties of materials are specified for the benchmark problem. Results from the first part will be used in code-to-code comparison. In the second part, the SAM model is used to simulate Test PG-27, which is the first PCC test carried out in the HTTF, with only two of the ten heater banks activated. The results in the second part are used for code-to-data comparison. Because the helium coolant flow rate is not measured in this facility, it is estimated using the input power and inlet/outlet coolant temperatures. Additionally, radial heat flow in the ceramic blocks is complicated by hundreds of cylindrical coolant channels and heater rods embedded in them. As such, it is necessary to deduce an effective thermal conductivity for the ceramic to analyze the core thermal behavior. SAM predictions of the helium coolant and ceramic temperatures are compared with test data measured in three equivalent sectors. Overall, the SAM results agree reasonably well with test data within the variation of data among the three sectors, which demonstrates SAM’s capability in capturing transient effects in HTGR using the simplified 2D ring model.

Improved multiphysics model of the High Temperature Engineering Test Reactor for the simulation of loss-of-forced-cooling experiments

We present a multiphysics model of the High Temperature Engineering Test Reactor for comparison with past and predict future loss-of-forced-cooling (LOFC) experiments. The approach selected combines (1) 3-D full-core superhomogenization-corrected neutronics, (2) 3-D full-core homogenized or semi-heterogeneous heat transfer (macroscale), (3) 2-D axisymmetric fuel rod heat transfer (pin-scale), and (4) 1-D thermal-hydraulics channels. Although large uncertainties remain, the time and magnitude of the first power peak after re-criticality is predicted within 1.5 h and 175 kW, respectively. The novelty of our work includes (1) a new macroscale/pin-scale heat transfer coupling approach relying on gap conductance to drastically speed up numerical convergence by two orders of magnitude, (2) determination of a radial effective thermal conductivity, reproducing the semi-heterogeneous re-criticality time within one hour using a homogenized macroscale model, and (3) a preliminary study of the reactor’s early behavior following a LOFC event, enabling further assessment of numerical models against fission power measurements.

Development and validation of multiphysics PWR core simulator KANT

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.

CorTAF: A nuclear reactor core three-dimensional thermal-hydraulic characteristics analysis code based on OpenFOAM

In this paper, a thermal hydraulic analysis method based on open source CFD platform OpenFOAM was proposed. The models of coolant flow and heat transfer, fuel rod heat conduction and coupled heat transfer were established according to the bundle structure characteristics of PWR core, and then a nuclear reactor three-dimensional thermal–hydraulic characteristics analysis code CorTAF was developed based on finite volume method with the spatialresolution of subchannel level. The international benchmarks, including GE3 × 3, Weiss and PNL2 × 6 fuel assembly flow and heat transfer experiments, were selected to carry out the validation. The calculation results were basically consistent with the experimental data, illustrating that CorTAF is suitable for predicting the flow and heat transfer characteristics of coolant in the rod bundle fuel assembly. The thermal–hydraulic characteristics of fuel assembly and PWR core under full power operation and blockage accident conditions were simulated using CorTAF. The steady-state and transient spatial distribution of physical quantities in subchannel-scale including coolant and fuel rod temperature were obtained, and the influence of blockage condition on flow and heat transfer characteristics was analyzed. This work has reference significance for the development of core thermal hydraulic analysis tools for PWR.

High Temperature Test Facility sensitivity and calibration studies to inform OECD-NEA benchmark calculations

The High Temperature Test Facility (HTTF) is a thermal hydraulics test facility for block-type high-temperature gas-cooled reactors (HTGRs) built to provide data to validate HTGR modeling tools. Several experiments were conducted at HTTF in the spring of 2019 that serve as the basis for an upcoming Organization for Economic Cooperation and Development – Nuclear Energy Agency HTGR thermal hydraulics benchmark. We present sensitivity studies of steady state and both the pressurized and depressurized conduction cooldown (PCC and DCC) transients in HTTF in this paper. These studies identified that uncertainties in the thermal conductivity of the ceramic blocks making up the HTTF core are significant drivers of peak temperatures in steady state, PCC, and DCC. The emissivity of the heater rods and core blocks also provide significant control over heater rod temperatures. We also performed calibration studies based on HTTF Experiment PG-26, which was intended to serve as the basis for the DCC benchmark problem. We found that no calibration was able to reproduce PG-26 peak block temperatures within an average of 30%, but by taking the best-estimate PG-26 calibration and applying it to a similar experiment—PG-29—we were able to reproduce the peak block temperatures within an average of 6.2% error. Based on this nearly fivefold improvement in peak block temperature predictions from PG-26 to PG-29 and confounding factors in PG-26 that contributed to previous difficulties modeling the experiment in RELAP5-3D and SAM we conclude that PG-26 data are insufficient for developing a DCC benchmark and recommend using PG-29 as the benchmark experiment instead.

BETHSY 9.1b test calculation with TRACE using 3D vessel component

Recently, several advanced multidimensional computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. One of such advanced, best-estimate reactor systems codes is TRAC/RELAP Advanced Computational Engine (TRACE), developed by the U.S. Nuclear Regulatory Commission. The advanced TRACE comes with a graphical user interface called SNAP (Symbolic Nuclear Analysis Package). It is intended for pre- and post-processing, running codes, RELAP5 to TRACE input deck conversion, input deck database generation etc. The TRACE code is still not fully development and it will have all the capabilities of RELAP5. The purpose of the present study was therefore to assess the 3D capability of the TRACE on BETHSY 9.1b test. The TRACE input deck was semi-converted (using SNAP and manual corrections) from the RELAP5 input deck. The 3D fluid dynamics within reactor vessel was modeled and compared to 1D fluid dynamics. The TRACE 3D calculation was compared both to TRACE 1D calculation and RELAP5 calculation. Namely, the geometry used in TRACE is basically the same, what gives very good basis for the comparison of the codes. The only exception is 3D reactor vessel model in case of TRACE 3D calculation. The TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used for calculations. The BETHSY 9.1b test (International Standard Problem no. 27 or ISP-27) was 5.08 cm equivalent diameter cold leg break without high pressure safety injection and with delayed ultimate procedure. BETHSY facility was a 3-loop replica of a 900 MWe FRAMATOME pressurized water reactor. In general, all presented code calculations were in good agreement with the BETHSY 9.1b test. The TRACE 1D calculation results are comparable to RELAP5 calculated results. For some parameters they are better, this is mostly due to better tuning of the break flow, what influences timing of the transient. When comparing TRACE 1D and TRACE 3D calculation, the latter is slightly better. One reason for comparable results is already good agreement of 1D calculations and there was not much space to further improve the results. The other reason may be that in the facility the phenomena were mostly one dimensional (for example, external downcomer was used for reactor vessel modeling). However, when 3D behavior of the heater rod temperatures was investigated, the advantage of three dimensional treatment was clearly demonstrated.

Prediction of reflood behavior in a bundle with deformed and un-deformed rods in ATHER 5 × 5 experiment using MARS-KS code

A reflood experiment at ATHER 5×5 rod bundle facility having an array of deformed rods is calculated using MARS-KS code. In the experiment, the deformed rods were represented by the heater rods with sleeves on the outside. To represent properly the configuration, the sleeve thermal conductors, the steam gap between the heater rod and the sleeve, and the fouling factor into the heat transfer area by contact between adjacent sleeves are implemented. The portion into which the sleeve is installed and the portion not installed are modeled as separate heat structures. Also the heat loss through the housing of the test section is modeled. Within the range of the present calculations, imposing a convective heat transfer coefficient of 2 W/m2·K over the housing can be considered appropriate for the actual heat loss. Calculations are conducted using one-dimensional model with single channel, one with two channels, and three-dimensional model by ‘multid’ component of the MARS-KS code. From the calculation results, the predictability of the code and the present modeling scheme is discussed in terms of overall response, peak cladding temperature, and time of quenching. Sensitivity study on interfacial friction model multiplier is also performed, which shows the quenching time can be improved by applying a multiplier 1.05 to the interface friction model in the one-dimensional case, but its effect on the maximum cladding temperature and the rewet temperature seems not to be significant. The differences in phenomena between the deformed rods and un-deformed rods and their reasons are discussed. The effect of 90% of partial blockage is expected less than 30 K in terms of maximum cladding temperature for both deformed and non-deformed rods, compared to the hypothetical case of the completely un-deformed rod bundle, because cooling can be maintained by flows through the remaining unblocked flow paths.

Experiments and CFD simulations of the LBE loop in HYST: A new concept for irradiation experiments in a fast-reactor-like environment

The current work investigates the thermal-hydraulics of a HYbrid fast/thermal core configuration Subcritical Testbed (HYST), a new concept proposed for irradiation experiments in a fast-reactor-like environment. Both the experimental studies and CFD simulations are performed to assist the development of HYST system. The experiments characterize heat transfer and pressure drop in the fast core as well as the heat transfer to thermal core from the fast core, while the CFD simulations only focus on the fast core to reduce the computational cost. The Reynolds-Averaged Navier Stokes (RANS) simulations using ANSYS Fluent are benchmarked through a two-step strategy first by existing data in literature and then by the newly collected data in HYST. By comparing with the existing data, confidence is gained in mesh generation, wire-rod contact modeling, and selected turbulence and Prt models. The mesh independence study in simulating HYST using Fluent indicates that the mesh convergence can be reached. Investigated parameters include the LBE temperature and velocity magnitude in all subchannels, heater rod surface temperature, outlet LBE temperature, and pressure drop between inlet and outlet. The effects of different Prt models and non-uniform power distribution on HYST simulation results are also investigated. Finally, these RANS simulations of HYST are benchmarked by preliminary experimental data in both adiabatic and heated tests. The experimental data of LBE outlet temperature for heated tests compare well with the simulations; however, the pressure drops are underestimated by the correlations and CFD simulations. In the next development phase, more detailed experimental studies and CFD simulations will be performed to assist the design of HYST system.

Critical heat flux characteristics of flow boiling on a heater rod under inclined and rolling conditions

Offshore floating nuclear power plants have been developed to supply power to remote regions. Since they are operated in an ocean environment, the thermal–hydraulic phenomena in the reactor core may be different from those in the land-based reactors. However, the experimental study on critical heat flux (CHF) under ocean condition is limited and its mechanism is unclear. In this study, an experiment was performed to measure CHF in a single heater rod under inclined and rolling conditions. To create the required environment to simulate the conditions of inclination and rolling, a rolling platform was designed and constructed. It could provide an inclination of up to 45° and rolling of up to 45° with a minimum period of 6 s. In the CHF test loop, the experimental conditions were set up to correspond to pressurized water reactor conditions, using the fluid-to-fluid scaling method. The working fluid used was R134a. The test section was a concentric annular channel composed of a heater rod and an unheated tube housing. The result of the inclination experiment suggested that in the most cases, the CHF was enhanced when compared with the vertical condition. In a limited number of cases of low-pressure and high-flow conditions, the CHF was degraded as compared to the vertical condition. These variations were proportional to the inclination angle and inversely proportional to the mass flux. From the results of the rolling experiment, a precursor of CHF was observed with a large fluctuation in temperature before the CHF, and its trigger was related to the CHF value of the inclined case. Based on this observation, the overall relations between the vertical, inclined, and rolling CHF values were derived. It was concluded that the rolling motion weakened the CHF enhancement and degradation effects of the inclination.

Conceptual design of gas cartridge loop components for the Versatile Test Reactor

In continuation of the Gas Cartridge Loop (GCL) development for the Versatile Test Reactor (VTR), General Atomics-Electromagnetic Systems (GA-EMS) conducted conceptual design studies of the GCL major components such as test articles, electromagnetic (EM) motor, heating elements, instrumentation and control (I&C), and pressure vessel. While the pre-conceptual design used a single test article and estimated the heating and cooling requirements of the GCL, the conceptual design used three test articles, with a reduced height by a factor of two, to enhance better experimental capabilities. The thermal-fluid analysis determined the performance requirements of the motor and heater, and confirmed the gas-cooled fast reactor (GFR) experimental conditions are achieved with the total test article power of 16.5 kW compared with 11 kW of the pre-conceptual design.The conceptual design of the motor is a four-pole three-phase high temperature induction motor. The motor and heater were designed with a 30% margin to the required power. Investigation was conducted on different heating element technologies, such as Nichrome, Kanthal, MoSi2, and SiC, for their temperature and power density limits. The conceptual design of the heater focused on material selection for the metal sheath and the heater rod arrangements. I&C design elaborated and clarified the conceptual design by defining the control logic and required number of lines for the redundancy and availability of the sensors and detectors of the GCL. Mechanical design analysis of the GCL pressure vessel conducted during the pre-conceptual design is still valid for the design limits of the stress and strain.