Small Modular Reactors (SMRs) adopting passive mitigation strategies are currently the most promising technology for the near term deployment of nuclear power generation. Different SMRs designs are currently under development and are, in general, characterized by some common features with the current reactors and by other features typical of their designs. Therefore, though numerous code validation study against natural circulation (NC) have been performed for large scale reactors, further analyses are necessary to characterize the capability of codes against available experimental data representative of SMR phenomenology. Though different scaling methodologies have been developed, considering the complex geometry and phenomena of a NPP, in the design of scaled-down experimental facilities it is not possible to avoid distortions, which should be limited to non-dominant phenomena. Even if dominant phenomena are preserved, due to the missing data at NPP scale, the code accuracy should be tested at different scales. Therefore, in a verification and validation process, the uncertainty related to the code scaling-up capability should be addressed, for example using counter-part tests. Since NC tests at different scales in integral test facilities devoted to SMR are currently not available, a numerical scaling methodology is here proposed. Based on previous activities, having as a reference the NC DOE tests developed in the OSU-MASLWR facility, the USNRC best estimate thermal hydraulic TRACE code has been validated for simulating NC in steady and transient conditions. Since the OSU-MASLWR is volume and height scaled, the target of this paper is to assess the scaling-up capability of the OSU-MASLWR Reactor Pressure Vessel nodalization in the prediction at different scales of NC and other phenomena typical of SMR, having as a base the OSU-MASLWR-002 single phase NC data. This, also, gives some first insights about the TRACE scaling-up capability against single-phase NC in integral type configurations.

This paper provides a comprehensive overview of the ongoing research endeavors conducted at Oregon State University School of Nuclear Science and Engineering (OSU NSE) in the field of Light-Water, Small, Modular Reactors (LW-SMR). Included in this paper is a thorough exposition of the integral testing activities carried out at the Multi-Application, Small, Light-Water, Reactor (MASLWR) facility, including both the methodologies employed and the outcomes obtained. The test facility can operate at pressure of 11.4 MPa and a temperature of 590 K. It serves as an integral test facility, with all components designed at a 1/3 scale height and a 1/254.7 volume scale in reference to the NuScale module. The facility operates in compliance with the NQA-1 program and has been instrumental in supporting NuScale's licensing process and the validation of TH (thermal-hydraulic) codes. This paper further expounds on the present operational status of the facility and delineates upcoming plans. Beyond the experimental capabilities, additional attention is directed towards an explication of the research capabilities intrinsic to the Nuscale Simulator control room, housed within the OSU Energy Exploration (E2) Center. Finally, the paper delves into an exploration of ongoing research concerning the Dynamical System Scaling (DSS) of Small Modular Reactors (SMRs). This paper serves as a consolidated resource offering insight into the multifaceted aspects of LW-SMR research conducted at Oregon State University, encompassing integral testing, facility status, control room research capabilities, and the DSS endeavors related to SMRs.

The development of the Modular High-Temperature Gas-Cooled Reactor is a significant milestone in advanced nuclear reactor technology. One of the concerns for the reactor’s safe operation is the effects of a loss-of-flow accident (LOFA) where the coolant circulators are tripped, and forced coolant flow through the core is lost. Depending on the steam generator placement, loop or intracore natural circulation develops to help transfer heat from the core to the reactor cavity, cooling system. This paper investigates the fundamental physical phenomena associated with intracore coolant natural circulation flow in a one-sixth Computational Fluid Dynamics (CFD) model of the Oregon State University High Temperature Test Facility (OSU HTTF) following a loss-of-flow accident transient. This study employs conjugate heat transfer and steady-state flow along with an SST k-ω turbulence model to characterize the phenomenon of core channel-to-channel natural convection. Previous studies have revealed the importance of complex flow distribution in the inlet and outlet plenums with the potential to generate hot coolant jets. For this reason, complete upper and lower plenum volumes are included in the analyzed computational domain. CFD results also include parametric studies performed for a mesh sensitivity analysis, generated using the STAR-CCM+ software. The resulting channel axial velocities and flow directions support the test facility scaling analysis and similarity group distortions calculation.

In the last decades, taking into account the operational experience of fission nuclear reactors, the nuclear international technical community started the development of advanced reactor designs in order to satisfy the demand of the people to improve the safety of NPPs and to take into consideration the industry needs to improve the economic efficiency and reduce the capital costs of nuclear power technology. In this framework Small Modular Reactors (SMR) adopting light water as coolant, taking advantage of the experience developed in current larger scale LWR, can bring advantages in terms of increasing “inherent safety” due to the integral type configuration, lower nominal power, and the adoption of passive mitigation strategy. Starting from some of the research activities on code applications developed by the authors along the last decade, and available in the public scientific literature, the target of this paper is to give some insights and recommendations for the future development of new research activities, in national and international frameworks, in relation to thermal hydraulics of SMRs.

Among the Next-Generation Nuclear Plant (NGNP) designs, the High-Temperature Gas-Cooled Reactors (HTGRs) are very attractive, due to their inherent safety features, high power conversion efficiency, and potential of providing high-temperature process heat. To perform a thorough safety study and to license these types of reactors, sufficient information needs to be provided about the phenomena that occur during accident scenarios. While several experimental research efforts have been dedicated in the past to investigate accident scenarios, knowledge gaps still exist in the phenomena characteristic of pressurized and depressurized conduction cooldown (PCC/DCC) transients as well as for normal operation scenarios. This paper summarizes the Oregon State University High Temperature Test Facility (HTTF) test matrix, experimental campaign, and selected tests results. High Temperature Test Facility is a scaled Integral Test Facility (IET) that is capable of mimicking scaled dimensions and operational conditions of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The goal of the High Temperature Test Facility is to provide experimental data on the DCC, PCC and normal operating scenarios of the reference Modular High-Temperature Gas Cooled Reactor design. The DCC, PCC, mixing, heat up and cooldown tests described in this paper were performed at prototypical Modular High-Temperature Gas Cooled Reactor temperatures, scaled initial pressure conditions (∼200 kPa), and thermal power input of less than 70 kW. Presented test data show temperature distributions in the High Temperature Test Facility core, upper plenum, cross duct, or lower plenum. Based on these temperature profiles attempts to investigate stratified flow, natural convection flow, heat up, cooldown and mixing phenomena are made. Furthermore, this paper evaluates the performed test campaign in the light of the Very High Temperature Gas-cooled Reactor Phenomena Identification and Ranking Table (PIRT) and proposes experiments to complement the existing PCC/DCC testing database for the validation of the thermal-hydraulic codes.

This paper investigates the fundamental physical phenomena associated with internal coolant flow distribution in the upper plenum of a prismatic Very High Temperature Reactor (VHTR) during normal operation. Previous studies have revealed the importance of complex flow distribution in the inlet plenum with the potential to generate low velocity or stagnation zones that can subsequently lead to the formation of hot spots in the reactor core and hot streaks in the lower plenum. It is therefore of interest to ensure that coolant is evenly distributed when entering multiple reactor coolant channels. Non-uniformity in the flow distribution is assessed for the reference Oregon State University High Temperature Test Facility (HTTF) case. The HTTF is a reduced scale model (1:4 in height and diameter) of the Modular High Temperature Gas-Cooled Reactor (MHTGR). The developed CFD model reflects complete and detailed geometry of the HTTF upper plenum along with upcomer and metallic core support structure (MCSS). CFD results include parametric studies performed for circulator mass flow rate variation and mesh sensitivity analysis, generated using Siemens Starccm+ software. Outcomes also include variable pressure and temperature boundary conditions impact on the flow distribution.

Molecular tagging velocimetry (MTV) is a nonintrusive velocimetry technique based on laser spectroscopy. It is particularly effective in challenging gas flow conditions encountered in thermal hydraulics where particle-based methods such as particle image (or tracking) velocimetry do not perform well. The main principles for designing and operating this diagnostic are presented as well as a set of gases that have been identified as potential seeds. Two gases [H2O and nitrous oxide (N2O)] have been characterized extensively for thermodynamic conditions ranging from standard temperature and pressure to environments encountered in integral effects test (IET) facilities for high-temperature gas reactors. A flexible, modular, and transportable laser system has been designed and demonstrated with H2O and N2O seed gases. The laser system enables determining the optimum excitation wavelength, tracer concentration, and timing parameters. Velocity precision and thermodynamic domain of applicability are discussed for both tracers. The spectroscopic nature of the diagnostics enables one to perform first-principle uncertainty analysis, which makes it attractive for validating numerical models. Molecular tagging velocimetry is demonstrated for two flows. First, in blowdown tests with H2O seed, the unique laser system enables one of the largest dynamic ranges reported to date for velocimetry: 5000:1 (74 dB). N2O-MTV is then deployed in situ in an IET facility, i.e., the High-Temperature Test Facility at Oregon State University, during a depressurized conduction cooldown (DCC) event. Data enable researchers to gain insights into flow instabilities present during DCC. Thus, MTV shows a strong potential to gain a fundamental understanding of gas flows in nuclear thermal hydraulics and to provide validation data for numerical solvers.

The High Temperature Test Facility (HTTF) is a quarter-scale integral-effect facility designed to study pressurized and depressurized conduction cooldowns (PCC and DCC, respectively) in high temperature, gas-cooled reactors (HTGR). This study focuses on DCC and aims to characterize the gas exchange between the reactor core and the reactor cavity to assess the risk for air ingress and subsequent potential for natural circulation core cooldowns. High-resolution velocity profile measurements are performed at the coolant pipe break location over a long time-scale using molecular tagging velocimetry. This non-intrusive, laser-based technique is applied for the first time in the field of nuclear thermal hydraulics, providing new insights for understanding the underlying phenomena at play during a DCC event. Such data are also valuable for validation of numerical simulations such as Reactor Excursion and Leak Analysis Program (RELAP). Results are reported here for various gas mixtures of helium (the coolant) and nitrogen (surrogate for air) to study the effect of density ratio. The short transient (30 s) lock-exchange phase is well captured with helium flow velocity between 0.8 and 2 m/s. The flow is shown to persist over much longer time-scales (hours) at a velocity of about 0.2 m/s, a likely consequence of gas mixing in the hot leg and in the lower plenum of the reactor. Flow visualization at the exit of the hot leg shows shear instabilities, which supports the hypothesis of significant mixing.

Multiphase Reactors Engineering and Applications Laboratory (mReal) has designed and constructed a scaled-down dual-channel facility to investigate plenum-to-plenum natural circulation heat transfer through two channels for coolant flow that would be encountered during a loss of flow accident in the prismatic modular reactor (PMR). Heat transfer characterization of the current facility has been investigated under different upper plenum and cooled channel outer surface temperatures using sophisticated flush mounted heat transfer sensors. Results show a reduction in the values of local heat-transfer coefficient and Nusselt number along the heated channel with increasing outer surface temperatures. One significant observation was the heat transfer reversal close to heated channel exit, where the heat flows from gas to the channel wall. This flow reversal is attributed to recirculation at the heated channel exit to the upper plenum. The average heat transfer results, when compared with previous literature, showed a similar qualitative trend. © 2016 American Institute of Chemical Engineers AIChE J , 63: 387–396, 2017

A molecular tagging velocity (MTV) technique is developed to non-intrusively measure velocity in an integral effect test (IET) facility simulating a high-temperature helium-cooled nuclear reactor in accident scenarios. In these scenarios, the velocities are expected to be low, on the order of 1 m/s or less, which forces special requirements on the MTV tracer selection. Nitrous oxide (N2O) is identified as a suitable seed gas to generate NO tracers capable of probing the flow over a large range of pressure, temperature, and flow velocity. The performance of N2O-MTV is assessed in the laboratory at temperature and pressure ranging from 295 to 781 K and 1 to 3 atm. MTV signal improves with a temperature increase, but decreases with a pressure increase. Velocity precision down to 0.004 m/s is achieved with a probe time of 40 ms at ambient pressure and temperature. Measurement precision is limited by tracer diffusion, and absorption of the tag laser beam by the seed gas. Processing by cross-correlation of single-shot images with high signal-to-noise ratio reference images improves the precision by about 10% compared to traditional single-shot image correlations. The instrument is then deployed to the IET facility. Challenges associated with heat, vibrations, safety, beam delivery, and imaging are addressed in order to successfully operate this sensitive instrument in-situ. Data are presented for an isothermal depressurized conduction cooldown. Velocity profiles from MTV reveal a complex flow transient driven by buoyancy, diffusion, and instability taking place over short (< 1 s) and long (> 30 min) time scales at sub-meter per second speed. The precision of the in-situ results is estimated at 0.027, 0.0095, and 0.006 m/s for a probe time of 5, 15, and 35 ms, respectively.

The High Temperature Gas-cooled Reactor (HTGR) is one of the most mature Gen IV reactor concepts under development today. The High Temperature Test Facility (HTTF) at Oregon State University is a test facility that supports the R&D needs for HTGRs. This study focuses on the issue of helium mixing after the core section in the HTTF, the results of which are generally applicable in HTGRs. In the HTTF, hot helium jets at different temperatures are supposed to uniformly mix in the lower plenum (LP) chamber. However, the level of mixing is not sufficient to reduce the peak helium temperature before the hot jet impinges the LP structure, which can cause issues with structural materials and operational issues in the heat exchanger downstream.The maximum allowable temperature variation in the outlet duct connected to the lower plenum is defined as 40 K (+/- 20 K from the average temperature), while the CFD simulations of this study indicate that the reference design suffers temperature variations in the duct as high as 100 K. To solve this issue, the installation of mixing-enhancing structures within the outlet duct were proposed and analyzed using CFD modeling. We show that using either an optimized "Kwiat" structure (developed in this study) or a motionless mixer installed in the outlet duct, the temperature variations can be brought dramatically, with acceptable increases in pressure drop. The optimal solution appears to be to install double motionless mixers with long blades in the outlet duct, which brings the temperature variation into the acceptable range (from 100 K down to 18 K), with a resulting pressure drop increase in the HTTF loop of 0.73 kPa (6% of total pressure drop). (C)2016 Elsevier B.V. All rights reserved.

Understanding the phenomenon of oxidation in graphite at elevated temperatures is key for many nuclear engineering applications. This is due to graphite having extremely favorable material qualities for nuclear applications. Both analytical and empirical correlations have been suggested that describe graphite oxidation. These correlations require basic assumptions about the nature of graphite oxidation and divide graphite oxidation into four regimes, or areas where the rate of oxidation is significantly different and likely driven by different, or modified, processes. Experiments were conducted for the construction of the Oregon State University (OSU) High Temperature Test facility (HTTF) where these underlying assumptions could be re-examined. This was done to understand how the graphite in the Oregon State University test facility will degrade over time and eventually fail. From these experiments it was investigated, and concluded, that the OSU HTTF will be able to complete its currently scheduled tests without failure in the test facility due to graphite oxidation. In accordance with industry practices the oxidation behaviors of the graphite were investigated using thermos-gravimetric analysis (TGA). This analysis allows for the precise measurement of mass, temperature, mass change, temperature change, and gas volume flow. Simultaneously additional experiments were done involving a tube furnace that allowed for the macroscopic and qualitative inspection of the graphite. The data from these tests was found to support the previously understood underlying assumptions in regimes not previously assumed. The data was analyzed to provide opportunities for further research in investigating graphite oxidation while specific oxidation rates in two of the regimes are reported for Tokai G-348 nuclear grade graphite.

In the short term period the use of advanced Small Modular Reactor (SMR) is one of the most promising options for the deployment of nuclear technology. The validation and assessment of the best estimate thermal hydraulic system code TRACE against SMR thermal hydraulic phenomena is a novel effort. In this framework the use of the natural circulation database developed at the OSU-MASLWR test facility, simulating the MASLWR reactor prototype, is of interest for analyses of the TRACE code capability in predicting natural circulation and primary/containment coupled behavior in SMR. The target of this paper is to analyze the TRACE V5 capability for the simulation of natural circulation phenomena, at different primary and secondary side conditions, and to simulate the primary/containment coupling behavior, typical of the MASLWR design in Beyond Design-Basis Accidents (BDBA), by using a 3D TRACE model of the containment. The results of the calculated data show that the TRACE code is able to predict, from a quantitative point of view, the primary natural circulation mass flow rate, and that a 3D TRACE model of the containment is able to predict the main thermal hydraulic parameters, characterizing the primary/containment coupled thermal-hydraulic behavior.

The U.S. Department of Energy (DOE) program to develop the Very High Temperature Gas Cooled Reactor (VHTR) with helium coolant is a technological cornerstone for advanced applications which further expands to the safe use of nuclear energy. The air ingress into the reactor vessel following a VHTR depressurization is ranked as important in regard to core safety. In the case of depressurized conduction cooldown event (DCC), the first stage of the accident is a loss of helium with depressurization. This occurs until atmospheric conditions are reached. After depressurization, when pressure in the system equalizes, air enter the reactor vessel through the stratified flow. The last stage of air ingress mechanisms are molecular diffusion and natural convection of air and helium inside the reactor pressure vessel. Reactor graphite components will produce exothermic reactions in the presence of oxygen at high temperatures. There is a danger that it may cause a loss of core structural integrity via oxidation or surface corrosion. This paper focuses on the first stage of DLOFC and analyzes the difficulties connected with modeling the depressurization phenomena using CFD tools (Star ccm+). MATLAB code was written to model the depressurization phase and compare the results obtained from Starccm+.

Particle image velocimetry boundary layer measurements are presented for flows over convex surfaces, and subject to flow acceleration. These mechanisms are present in gas reactor core designs, which is of concern because it is known that they can cause boundary layer laminarization which can have a negative effect on surface heat transfer. A wind tunnel was constructed where flow acceleration and curvature effects were applied separately, and simultaneously to investigate the separate and combinative influence of these mechanisms. Applied acceleration for the different test cases varied from K = 1.6 x 10(-6) to 1 x 10(-5). Curvature values of delta/R= 0.015 to delta/R= 0.05 were tested. The applied acceleration spans across the threshold value when laminarization is expected to occur for flat plate geometries.Measurements were taken using particle image velocimetry. Mean flow parameters including mean velocity profiles, boundary layer thickness, and shape factor are presented as a function of streamwise position during the boundary layer response as it is subjected to the laminarization causing mechanisms. Velocity profiles are normalized by both outer and inner variables.It is observed that as the flow is subjected to these mechanisms, the boundary layer mean flow parameters diverge from their turbulent values and approach laminar characteristics. The linear viscous sublayer increases in thickness, the overall boundary layer thickness is reduced, and the shape factor increases. The response in the boundary layer mean flow parameters is more pronounced and occurs more quickly when subject to both mechanisms simultaneously. (C) 2014 Elsevier B.V. All rights reserved.

The integral Pressurized Water Reactor (PWR) concept, which contains the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high possibility for near-term deployment. An IAEA International Collaborative Standard Problem (ICSP) on “Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Primary System and Containment during Accidents” has been conducted since 2010. Oregon State University of USA has offered their experimental facility, which was built to demonstrate the feasibility of Multi-Application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven IAEA Member States have been participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiment. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena including the coupling of primary system, high pressure containment and cooling pool are expected to occur in this transient. The ICSP has been conducted in three phases: pre-test (with designed initial & boundary conditions before the conduction of the experiment), blind (with real initial & boundary conditions after the conduction of the experiment) and open simulation (after the observation of real experimental data). Most advanced thermal-hydraulic system analysis codes like TRACE, RELAP5-3D and MARS have been assessed against experiments conducted at MASLWR test facility.

In support of the conversion of the Oregon State TRIGA Reactor (OSTR) from highly enriched uranium (HE U) fuel to low-enriched uranium (LEU) fuel, a comprehensive neutronic analysis utilizing MCNP5 was performed on the HEU and LEU core configurations. The initial 1974 HEU core provided an opportunity for verification of the MCNP5 baseline model; all fuel elements in the initial core were congruent in geometry and material composition, having no burnup. In addition, a substantial database of core parameters was documented during the initial HEU core start-up. This verification study examined control rod worth, core excess reactivity, burnup, core power, power per element, temperature coefficient of reactivity, void coefficient of reactivity, moderator coefficient of reactivity, axial and radial power profiles, prompt-neutron lifetime, effective delayed-neutron fraction, power defect, and xenon poisoning.Fuel material composition and core loadings are presented. The excellent comparison between the numerical results and the experimental data of the initial HEU core established an objective, credible baseline model and methodology, which were then extended to the LEU core neutronic analysis. Comparison between the numerically calculated core physics values for the new LEU core and data collected during start-up provided a complete verification that the MCNP5 models developed for both the HEU and LEU cores were representative of the OSTR.