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GAMMA Multidimensional Multicomponent Mixture Analysis to Predict Air Ingress Phenomena in an HTGR
Nuclear Science and Engineering
Abstract
We developed a multidimensional G As Multicomponent Mixture Analysis (GAMMA) code in order to investigate chemical reaction behaviors related to an air ingress accident and the thermofluid transients in high-temperature gas-cooled reactors. The implicit continuous Eulerian technique is adopted for the reduction of a 10N × 10N matrix into an N × N pressure difference matrix and fast transient computation. In the validation with a high-temperature engineering test reactor (HTTR)-simulated air ingress experiment, the onset times of natural convection are accurately predicted within a 10% deviation. Small internal leaks in the HTTR-simulated test facility have been found to significantly affect the consequence of air ingress. In all the simulated cases for a SANA-1 afterheat removal test, the predictions of GAMMA are in a high level of agreement with the measured temperature profiles and are comparable to the results of other codes (TINTE, THERMIX/DIREKT, and TRIO-EF).
... In the axial direction, the porosity is a damped, oscillatory function of the distance from bounding walls with smooth transitions to unity near free surfaces that depend on the weight of the pebbles [120,121,123,125,144]. However, most simulations of PBRs neglect the axial dependence of porosity, likely due to the lack of models for the axial distribution [101,104,108,[145][146][147][148][149]. An approximate method for extending radial porosity distributions to include axial dependence is developed later in this section. ...
... Within the nuclear engineering field, several non-commercial tools exist for modeling PBR T/H. Examples of these tools include the German Forschungszentrum Jülich Research Centre THERMIX application, widely used in the design and analysis of PBRs in Germany and South Africa [65,101,215,216]; the University of Michigan and Nuclear Regulatory Commission (NRC) Advanced Gas REactor Evaluator (AGREE) application frequently applied to prismatic gas reactor analysis [206]; the Korean Atomic Energy Research Institute (KAERI) GAMMA application [147]; the Rensselaer Polytechnic Institute PEBble Fluid Dynamics (PEBFD) application [125]; and the Iranian Sharif University of Technology Thermo Hydraulic Porous Program (THPP) application [108]. ...
... Section 5.3 then presents simulation results for all 52 of the steady-state and axisymmetric experiments performed, plus an open upper plenum case, using a single set of baseline macroscale closures. When available, code-to-code comparisons are made to Flownex and GAMMA, two porous media applications that have previously been used to model gas-cooled PBRs [146,147]. Simulation data has graciously been provided by Dr. C. G. Du Toit of the School of Mechanical and Nuclear Engineering at North-West University, South Africa and Dr. H. S. Lim of KAERI. ...
Pebble bed reactors (PBRs) are expected to display excellent heat removal characteristics due to graphite's capability for storing and transferring heat, the high failure temperatures of particle fuel, and the low power densities involved. However, a major challenge associated with the modeling of PBRs is the complex fuel-coolant structure in the core. Thermal-hydraulic (T/H) modeling of PBRs requires consideration of thermal and flow effects over five orders in spatial magnitude - from 5e9 fuel particles, each about 1 mm in size, to 5e5 pebbles, each about 5 cm in size, within a 10 m-size reactor core in the larger context of a power generating system. This research develops and applies multiscale methods to the thermal analysis of PBRs. By decomposing the complex PBR geometry into coupled models for three characteristic length scales - the particle, pebble, and core - efficient predictions of core T/H relevant to reactor design are achieved.
These multiscale models are implemented in a new finite element software application built on the open source Multiphysics Object-Oriented Software Environment (MOOSE). By leveraging state-of-the-art numerical methods, solvers, and meshing tools, this dissertation enables rapid design and analysis for scoping studies, fast-turnaround design, and multiphysics coupling to a comprehensive reactor analysis framework.
Application of multiscale analysis to a wide variety of flows demonstrates the software tool's capabilities as a general flow solver and the applicability of these models to both porous and open flows. Verification for open flows shows that the multiscale model reduces to the Navier-Stokes equations in regions such as reactor plena, where prediction of mixing and the ensuing thermal stresses are essential to the design of reactor internals. Application to the SANA experiments, a gas-cooled scaled PBR facility, demonstrates that the multiscale model predicts the passive conduction cool-down heat removal process with an average solid temperature error of 22.6 C. Statistical analysis as a function of position within the bed and other experimental characteristics highlights limitations of model closures and simplifications that are useful in guiding further macroscale analysis of gas-cooled PBRs.
Supported by the verification and validation for open flows and gas systems, full-core steady-state T/H analysis is performed for the Mark-1 Pebble Bed Fluoride-Salt-Cooled High-Temperature Reactor (PB-FHR). The unconventional reflector block design, uniquely thin fuel-matrix annulus, and non-uniform flow boundary conditions (BCs) highlight the new capabilities enabled by this research for PBR industrial analysis. Two multiscale fuel models are compared against full-resolved PB-FHR fuel pebbles for a wide range in thermal conditions. While a homogeneous layer model is characterized by errors in excess of 200 C, a linear superposition method is shown to predict average and maximum temperatures to within 10 C.
Early models of PBRs have struggled to accurately characterize the core bypass fraction, with significant implications on fuel temperature predictions. A porous media model is constructed of the reflectors corresponding to the maximum-bypass end-of-life condition with friction factor correlations generated using COMSOL Multiphysics. A tensor representation of the friction factor shows that the momentum loss is significantly higher in the radial than the axial direction. These drag models are combined with the multiscale fuel verification for full-core analysis of the PB-FHR. A parametric study varying the reflector block gap distribution and the inflow port design demonstrates that the inflow BC has a significant effect on the core bypass fraction and that the bypass fraction is a strong function of the reflector block gap distribution. The maximum bypass fraction is predicted to be within the range of 11.9%-14.0% depending on the inflow BC and gap distribution. For a bottom-heavy center reflector inlet, fuel and reflector temperature predictions are provided. The primary effect of the core bypass is to uniformly raise core temperatures; for all reflector gap distributions, the maximum kernel temperature is approximately 93 C higher than the maximum fluid temperature, which remains far below the fuel failure limit.
This work demonstrates the utility of multiscale methods to thermal analysis of PBRs. In conjunction with the larger scientific community, this research enables fast-turnaround design and analysis of all single-phase PBRs to facilitate the contribution of advanced nuclear reactors to a clean energy future.
... For instance, while Pronghorn uses a 2-D porosity distribution, Flownex and GAMMA both neglect any axial dependence. 44,45 In addition, the κ s model used in Flownex and GAMMA does not account for wall effects, while the Tsotsas radiation correction is used in the Pronghorn model. 44,45 And while all three codes assume the same pebble emissivity, differences exist in the calculation of the vessel surface convection coefficient. ...
... 44,45 In addition, the κ s model used in Flownex and GAMMA does not account for wall effects, while the Tsotsas radiation correction is used in the Pronghorn model. 44,45 And while all three codes assume the same pebble emissivity, differences exist in the calculation of the vessel surface convection coefficient. Based on the median assumed by other benchmark participants, a fixed vessel surface convection coefficient of 15 W m −2 K −1 was assumed in Pronghorn. ...
This paper presents an overview of Pronghorn, a multiscale thermal-hydraulic (T/H) application developed by Idaho National Laboratory and the University of California, Berkeley. Pronghorn, built on the open-source finite element Multiphysics Object-Oriented Simulation Environment (MOOSE), leverages state-of-the-art physical models, numerical methods, and nonlinear solvers to deliver fast-running advanced reactor T/H simulation capabilities within a modern software engineering environment. This work summarizes the physical models, multiphysics and multiscale coupling, and numerical discretization in Pronghorn with emphasis on our initial target application to pebble bed reactors (PBRs). A diverse set of applications are shown to depressurized natural circulation in the SANA experiments, forced convection in the Pebble Bed Modular Reactor, three-dimensional (3-D)/one-dimensional coupling of Pronghorn and RELAP-7 systems T/H for loop analysis in the High Temperature Reactor Power Module, and forced convection in the Mark-1 Pebble Bed Fluoride-Salt-Cooled High-Temperature Reactor. A multiphysics coupling of Pronghorn, RELAP-7, and Griffin deterministic neutronics for a gas-cooled PBR demonstrates the capability of the MOOSE framework for reactor design calculations. These applications highlight the verification and validation underlying Pronghorn's software development while emphasizing features that improve upon capabilities offered by legacy tools in areas such as 3-D unstructured meshing, physics modeling, and multiphysics coupling.
... This study deals with the comparison of the results obtained with the GAMMA+ [10] and Flownex [11] codes of the expected flow and heat transfer in the RCCS. The benchmark calculation estimates total heat loss, radiation and natural convection heat transfer from the RPV to the RCCS at different fixed temperatures on the inside surface of the RPV. ...
... This study dealt with the comparison of the results obtained with GAMMA+ [10] and Flownex [11] of the expected flow and heat transfer in the RCCS of the MW th PMR (PMR200) General Atomic prismatic core VHTR reactor under consideration by Korea [9]. The benchmark calculation estimated total heat loss, radiation and natural convection heat transfer from the RPV to the RCCS at different fixed temperatures on the inside surface of the RPV. ...
... The other available option at KAERI could be the use of system codes such as MARS-GCR [16] and GAM-MA+ [17]. It should be noted, however, that the system codes are targeted for system transients. ...
... With the known field of δP i , δQ j can be obtained using Eq. (17). Then, new values for P i , Q j , and ρ f,i are determined using Eqs. ...
A new computer code, named CORONA (Core Reliable Optimization and thermo-fluid Network Analysis), was developed for the core thermo-fluid analysis of a prismatic gas cooled reactor. The CORONA code is targeted for whole-core thermo-fluid analysis of a prismatic gas cooled reactor, with fast computation and reasonable accuracy. In order to achieve this target, the development of CORONA focused on (1) an efficient numerical method, (2) efficient grid generation, and (3) parallel computation. The key idea for the efficient numerical method of CORONA is to solve a three-dimensional solid heat conduction equation combined with one-dimensional fluid flow network equations. The typical difficulties in generating computational grids for a whole core analysis were overcome by using a basic unit cell concept. A fast calculation was finally achieved by a block-wise parallel computation method. The objective of the present paper is to summarize the motivation and strategy, numerical approaches, verification and validation, parallel computation, and perspective of the CORONA code.
... Such an approach offers little insight into the kinetics of the chemical reactions taking place and could not be used to calculate the production rates of CO and CO 2 gases nor their relative contributions to the total gasification. In addition, the large variances in reported values by different investigators of the apparent activation energy and pre-exponential coefficient in the Arrhenius rate relations, even for the same nuclear graphite grade, result in a wide range of gasification rate predictions Xiaowei et al., 2004;Yu et al., 2008;Takeda, 2004;Lim and No, 2006). And since the prevailing gasification mode of the porous nuclear graphite strongly depends on temperature, the values of the apparent activation energy and preexponential coefficients are different in different temperature ranges. ...
... In the diffusion-limited mode (c) of nuclear graphite gasification at high temperatures (Fig. 1), other investigators (Takeda, 2004;Lim and No, 2006;Kakaç and Yener, 1995;Kim and No, 2006) used a Graetz solution (Kakaç and Yener, 1995) that is based on the similarity of heat and mass transfer to calculate Sh and hence, the oxygen diffusion velocity, k m through the surface boundary layer. The Graetz solution has been developed for laminar gas flow through uniformly heated tubes. ...
Gasification of nuclear graphite in the unlikely event of massive air ingress in High-Temperature and Very-High Temperature gas-cooled Reactors is a safety concern, requiring accurate and reliable predictions of the erosion rate of the external surface and within volume pores. At low temperature, gasification occurs within the open pores gradually degrading the mechanical strength of graphite components. Gasification shifts gradually to the external surface with increasing temperature. At high temperatures, although the rates of chemical reactions increase exponentially with temperature, they are limited by the oxygen diffusion to the external surface. A semi-empirical Sh correlation is developed to calculate the oxygen diffusion velocity. It is based on an extensive database of reported measurements of the convective heat transfer coefficient for heated wires and cylinders in air, water and paraffin oil flows at 0.006 ≤ Re ≤ 2.42 × 105 and 0.068 ≤ Pr ≤ 35.2 and the mass transfer coefficient at 4.8 ≤ Re ≤ 104 and Sc = 0.609 and 1300–2000. The database also includes reported values of the averaged Sh for gasification of a cylinder of V483T nuclear grade graphite (300 mm long and 200 mm in dia.) at 1141–1393 K in ascending cross-flow of nitrogen gas containing 5 vol.% oxygen at 533 ≤ Re ≤ 1660. The Sh correlation is within ±8% of the compiled 807 data points and applicable to both internal and external parallel and cross-flow conditions. When implemented in a chemical-reaction kinetics model, the calculated gasification rates are consistent with reported measurements for different size specimens of nuclear graphite grades NBG-18, NBG-25, IG-11, IG-110 and IG-430 at intermediate and high temperatures in atmospheric air (0.08 ≤ Re ≤ 30).
... In addition, the large variances in the reported values of the apparent activation energy and pre-exponential coefficient by different investigators, even for the same grade of nuclear graphite, result in a wide range of predictions. [2][3][4][5][6] A challenging, but practical and consistent approach is using a chemical-reaction kinetics model that is fundamentally, rather than empirically based, to estimate the gasification rates of nuclear graphite. Such an approach has been developed and successfully validated with the reported measurements of gasification rates and weight loss in experiments. ...
... (7)) with the Graetz solution 14 , Eq. (6), used by Kim and No 13 for calculating the diffusion velocity to match their experimental measurements of the total gasification rate for IG-110 nuclear graphite at high temperatures (or Mode (c)). Figure 4 compares the correlations in Eqs. (6) and (7) for 0.001 < Re < 200. Unlike Eq. (6), indicating that for Re < 1.0, Sh is constant and = 3.66, the present Sh correlation (Eq. ...
The safety analysis of High-Temperature and Very High Temperature gas-cooled Reactors requires reliable estimates of nuclear graphite gasification as a function of temperature, among other parameters, in the unlikely event of an air ingress accident. Although the rates of prevailing chemical reactions increase exponentially with temperature, graphite gasification at high temperatures is limited by oxygen diffusion through the boundary layer. The effective diffusion velocity depends on the total flow rate and pressure of the bulk air-gas mixture. This paper developed a semi-empirical Sherwood number correlation for calculating the oxygen diffusion velocity. The correlation is based on a compiled database of the results of convective heat transfer experiments with wires and cylinders of different diameters in air, water and paraffin oil at 0.006 < Re < 1,604 and 0.068 < Sc < 35.2, and of mass transfer experiments at 4.8 < Re < 77 and 1,300 < Sc < 2,000. The developed correlation is within + 8% of the compiled database of 567 data points and consistent with reported gasification rate measurements at higher temperatures in experiments using different size specimens of nuclear graphite grades of NBG-18 and NB-25, IG-11, IG-110 and IG-430 in atmospheric air at 0.08 < Re < 30. Unlike the Graetz solution that gives a constant Sh of 3.66 at Re < 1.0, in the present correlation Sh decreases monotonically to much lower values with decreasing Re.
... Development of numerical methods to analyze thermofluid phenomena during normal and/or off-normal conditions of a High Temperature Gas-cooled Reactor (HTGR) has been an active research area all around the world for the past several years. Examples of the outstanding efforts are the GAMMA+ code [1][2][3] in Korea Atomic Energy Research Institute (KAERI) and the AGREE code [4][5] in the University of Michigan (U of M). The GAMMA+ code is targeted to support the NHDD (Nuclear Hydrogen Development and Demonstration) program in Korea. 1 It has the capability to calculate thermal-fluid transients as well as chemical reactions in a multi-component mixture system. ...
... The results of the validation study will then follow. More detailed descriptions of the codes are available in Ref. 3 ...
For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR), intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI) and the AGREE code of the University of Michigan (U of M). One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU) in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.
... Reactor analysis codes for HTGR, for example, DYN3D-HTR [1], PHISICS/RELAP5-3D [2], has developed during such researches. Korea Atomic Energy Research Institute (KAERI) has also developed key reactor core design codes such as reactor physics analysis code CAPP [3][4][5], high-fidelity thermal-fluid analysis code CORONA [6,7], and thermal-fluid/system safety analysis code GAMMA+ [8,9], which are targeted at block-type HTGRs. Through these, it is possible to perform the reactor core analysis and the safety analysis for block-type HTGRs. ...
Recently, the coupling between computer codes that simulate different physical phenomena has attracted for more accurate analysis. In the case of high-temperature gas-cooled reactor (HTGR), the coupling between neutronics and thermal-fluid analysis is necessary because of large change of temperature in the reactor core. Korea Atomic Energy Research Institute (KAERI) has developed the coupled code system between a reactor physics analysis code CAPP and a thermal-fluid system safety analysis code GAMMA+ for a block-type HTGR. The CAPP/GAMMA+ coupled code system provides more accurate block-wise distribution data than CAPP or GAMMA+ stand-alone analysis. However, the block-wise distribution data has the limitation in order to predict safety parameters such as the maximum temperature of the nuclear fuel. It is necessary to calculate refined distribution, for example, pin-level (fuel compact level) distribution. In this study, we tried to solve this problem by coupling CAPP and a high-fidelity thermal-fluid analysis code CORONA. CORONA can perform a high-fidelity thermal-fluid analysis of Computational Fluid Dynamics (CFD) level by dividing a block-type HTGR core into small lattices. On the other hand, CAPP can provide a pin power distribution. It is expected that the refined, more accurate distribution data for a block-type HTGR can be obtained by coupling these two codes. This paper presents the development of coupled code system between CAPP and CORONA, and then it is tested on a simple HTGR column problem with encouraging results.
... A fracture model of nuclear graphite was proposed by INL based on the structure tests of the oxidized core bottom structures. A multidimensional code named GAMMA was developed by Korea Atom Energy Research Institute (KAERI) to predict the air ingress accident of hightemperature gas-cooled reactor [8] and the advanced air ingress-related models developed by INL was implemented into the GAMMA code. Shiga discussed the process of air ingress during a depressurization accident of GTHTR300 [9]. ...
The double-ended guillotine break (DEGB) of the horizontal coaxial gas duct of a high-temperature gas-cooled reactor is an extremely hypothetical accident, which could cause the air to enter into the primary circuit and react with graphite in the reactor core. The performance of the HTR-PM plant under this extremely hypothetical accident has been studied by the system code TINTE in this work. The results show that the maximum fuel temperature will not reach the temperature design limitation, and the graphite oxidation will not cause unacceptable consequences even under some conservative assumptions. Moreover, nitrogen and helium injected from the fuel charging tube were studied as the possible mitigation measures to further alleviate the consequences of this air ingress accident. The preliminary results show that only the flow rate of nitrogen injected reaches a certain value, which can effectively alleviate the consequences, while for helium injection, both high and small flow rate can prevent or cut off the natural circulation and alleviate the consequences. The reason is that helium is much lighter than nitrogen, and the density difference between the coolant channel and the reactor core is small when helium is injected. Considering the injection velocity, the total usage amount, and the start time of gas injection, helium injected with a small flow rate is suggested.
... Table III summarizes the graphite oxidation models used in this calculation. These oxidation models are well validated by comparisons with the experimental data implemented into GAMMA code in the previous research (Lim and NO [13]). ...
An air-ingress accident in a VHTR is anticipated to cause severe changes of graphite density and mechanical strength by oxidation process resulting in many side effects. However, quantitative estimations have not been performed yet. In this study, the focus has been on the prediction of graphite density change and mechanical strength using a thermal hydraulic system analysis code. For analysis of the graphite density change, a simple graphite burn-off model was developed based on the similarity concept between parallel electrical circuit and graphite oxidation considering the overall changes of the graphite geometry and density. The developed model was implemented in the VHTR system analysis code, GAMMA, along with other comprehensive graphite oxidation models. GT-MHR 600 MWt reactor was selected as a reference reactor. From the calculation, it was observed that the main oxidation process was derived 5.5 days after the accident following natural convection. The core maximum temperature reached up to 1400 o C. However it never exceeded the maximum temperature criteria, 1600 o C. According to the calculation results, most of the oxidation occurs in the bottom reflector, so the exothermic heat generated by oxidation did not affect the core heat up. However, the oxidation process highly decreased the density of the bottom reflector making it vulnerable to mechanical stress. In fact, since the bottom reflector sustains the reactor core, the stress is highly concentrated on this part. The calculations were made for up to 11 days after the accident and 4.5% of density decrease was estimated resulting in 25% mechanical strength reduction.
... Numerical simulations performed by Oh et al. 11,20,21 using a FLUENT computational fluid dynamics code suggested that onc would occur within 100 s after the depressurization phase. However, the simulations performed by using a Gamma code 22,23 suggested that the onc would occur over several hours rather than several tens or hundreds of seconds after the depressurization event. Thus, our work is aimed at experimentally determining onc times for different reactor core temperatures and understanding the air transport phenomena in the reactor core following air ingress into the lower plenum. ...
In order to investigate air-ingress phenomena in a gas-cooled very high temperature reactor (VHTR), natural circulation experiments have been conducted in a helium flow loop after the injection of nitrogen into the lower plenum. A pair of helium analyzers were used to measure the nitrogen and helium concentrations in the lower plenum and upper plenum. The changes in the nitrogen concentration in the upper plenum were used to calculate the time required for the transport of nitrogen from the lower plenum to upper plenum through a riser flow channel made of graphite. The effect of system temperature and pressure on the rate of nitrogen transport has been studied extensively. Furthermore, a close examination of the graphite flow channel wall temperatures at different elevations showed small but sudden drops indicating the arrival of nitrogen at each elevation. From these data, the upward transport of nitrogen injected into the lower plenum under natural circulation conditions could be quantitatively investigated. The experimental findings indicate that the driving mechanisms for air transport through the reactor core of VHTR would result from both molecular diffusion and natural circulation. At low graphite temperatures in the riser, molecular diffusion is the dominating mechanism; however, as the riser temperature increases, natural circulation becomes dominant and the rate of nitrogen transport increases. Further, the time constants for these mechanisms have been calculated using a simplified species transport equation.
... The experiments and analytical examples are modeled using the codes, and simulation results obtained by the codes are then compared with the corresponding experimental and analytical results. Rousseau et al. (2015) concluded that for a complete RCCS system the systems codes GAMMA+ (Lim and No, 2006) and Flownex (M-Tech Industrial, 2015) have the ability to correctly solve the fundamental conservation equations and heat transfer relations in an integrated manner. The current paper is concerned with two benchmark problems for which the analytical solutions can be obtained and used to validate the heat transfer capability of systems codes, such as GAMMA+ and Flownex, to correctly calculate conduction, radiation, and convection, including mixed convection. ...
... The GAMMA+ code [30,31] is a system/safety analysis code for thermal-fluid and system transient, developed by KAERI. GAMMA+ enables the investigation of various phenomena occurring in a high-temperature gas-cooled reactor, such as pressurized or depressurized conduction cool down and air/ water-ingress accidents. ...
This paper presents the application results of MCS/GAMMA+ to multi-physics analysis of OECD/NEA modular high temperature gas-cooled reactor (MHTGR) benchmark Phase I Exercise 3. It is a part of international R&D efforts lead by the Next Generation Nuclear Plant (NGNP) US project to improve the neutron-physics and thermal-fluid simulation of (high temperature gas-cooled reactors) HTGRs, one of the next generations of safer nuclear reactors. Accurate and validated analysis tools are indeed a crucial requirement for safety analysis and licensing of nuclear reactors. To guide this effort, a numerical benchmark on the MHTGR was created by the NGNP project and formally approved in 2012 for international participation by the OECD/NEA. The benchmark defines a common set of exercises and the comparison of solutions obtained with different analysis tools is expected to improve the understanding of simulation methods for HTGRs. The coupled neutronics/thermal-fluid solution presented in this paper was obtained with the neutron transport Monte Carlo code MCS developed by Ulsan National Institute of Science and Technology and the thermal-fluid code GAMMA+ developed by Korean Atomic Energy Research Institute. The purpose of this paper is to present the GAMMA+/MCS coupled system, the calculation methodology, and the obtained solutions.
... JAERI designed a U shape test rig to study air diffusion and convection phenomena (Takeda, 2010). KAERI developed a multidimensional code to predict air ingress phenomena in the prismatic GT-MHR (Hong and Hee, 2006;Hee et al., 2007). MIT and INL simulated air ingress in the prismatic VHTRs with MELCOR and CFD (Kadak and Zhai, 2006;Richard et al., 2002). ...
... CFX and STAR-CD are commercial computational fluid dynamics codes. GAMMA+ is a VHTR system code that models the reactor on a scale similar to AGREE [78][79][80]. The tabulated results, shown in Table 2.3, show that the results from AGREE compare well with experiment. ...
In this thesis the various methods of obtaining sensitivity derivatives are explored in order to highlight the significant contributors of uncertainty, and the adjoint approach is shown to be the most efficient. The implementation of the
adjoint method into the NRC Advanced Gas REactor Evaluator (AGREE) code is performed. Additionally, a method for obtaining an approximation that has a leading error term of third order is derived by calculating both forward and adjoint sensitivities. The approximation is demonstrated to be accurate locally. A method for combining these surrogates is demonstrated using the DAKOTA code.
The utility of the method is demonstrated by performing three different analyses
related to the High Temperature Test Reactor (HTTR) in Japan. First, the bypass flow experiment of Kaburaki and Takizuka is analyzed to inspect the sensitivity and variability of the bypass flow with respect to the cross-flow gap geometry, loss coefficient and boundary conditions. Next, the HENDEL experiment is
analyzed to investigate the factors impacting peak core temperatures and the initial amount of stored energy in the core. Last, an analysis of the actual HTTR reactor is performed. A key result from these analyses is that the cross-flow gap geometry and loss coefficients were of relatively low significance with regard to
the quantities that pertain to core temperature.
... As a part of the Nuclear Hydrogen Development and Demonstration (NHDD) program in Korea, the Korea Atomic Energy Research Institute (KAERI) has been developing computer software to analyze the behaviors of the fission products (FP) circulating in the primary coolant loop and in the containment for VHTRs. This software, called GAMMA-FP (GAs Multi-component Mixture Analysis-Fission Products module), is being developed as an annex module of the previously developed thermal-fluidic analysis code, GAMMA+ (GAs Multi- [6]. In this way, the thermal-fluidic information required for the GAMMA-FP calculations such as temperature, pressure, and corresponding material properties at every transient time step can be directly utilized from the concurrent GAMMA+ calculation. ...
Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.
... They typically calculate the total gasification of nuclear graphite using semiempirical Arrhenius expressions. These expressions use constant apparent activation energies and pre-exponential rate coefficients determined from the best fit of total gasification rate measurements in laboratory experiments with small specimens; 0.25 to 50 g in total mass (Maruyama et al., 1995;Xiaowei et al., 2004;Yu et al., 2008;Kim et al., 2008;Takeda, 2004;Lim and No, 2006;No et al., 2007). The safety analyses use the Arrhenius gasification expressions, in conjunction with a control-volume or finite-elements numerical thermalhydraulics analysis of the reactor core. ...
This paper investigates the transient gasification of NBG-18 nuclear graphite with atmospheric air ingress in a 0.8-m long coolant channel of a prismatic Very High Temperature Reactor fuel element. Analysis varied the initial graphite and air inlet temperature, To, from 800 to 1100 K at air inlet Reynolds number, Rein = 5, 10 and 20. The analysis employs a Generic Interface that couples a multi-species diffusion and flow model to readout tables of the CO and CO2 production fluxes. These fluxes are functions of the graphite local surface temperature, oxygen partial pressure and graphite weight loss fraction and calculated using a chemical-reactions kinetics model for the gasification of nuclear graphite. The analysis accounts for the heats of formation of CO and CO2 gases, the heat conduction in the graphite sleeve, and the change in the oxygen partial pressure in the bulk gas flow mixture along the channel. Transient calculations performed up to a weight loss fraction of 0.10 at the entrance of the channel, t10. They include the local graphite surface temperature and composition of bulk gas flow, the local and total graphite weight losses and the local and total production rates of CO and CO2 gases. The heat released in the exothermic production reactions of these gases increases the local graphite surface temperature, accelerating its gasification. At the end of the calculated gasification transient, t = t10, the graphite weight loss is highest at the channel entrance and decreases rapidly with axial distance into the channel, to its lowest value where oxygen in the bulk gas flow is depleted. Increasing To decreases t10 and the total graphite loss, while increasing Rein decreases t10 but increases graphite loss.
... This is because of the low oxidants concentration, low oxidation rate, and uniform carbon gasification rate per unit of surface area of the open pores. A one-step power law has been used successfully to predict the reaction rate of graphite by air or oxygen in Mode (a) as: 18,28,35 ...
A massive air or steam ingress in High Temperature Reactors (HTRs) nominally operating at 600-950 o C is a design-basis accident requiring the development and validation of graphite oxidation and erosion models to examine the impact on the potential fission products release and the integrity of the graphite core and reflector blocks. Nuclear graphite is of many types with similarities but also differences in the microstructure, volume porosity, impurities, type and size of filler coke particles, graphitization and heat treatment temperatures, and the thermal and physical properties. These as well as the temperature, types and partial pressures of oxidants affects the prevailing oxidation mode and kinetics of the oxidation processes of graphite in HTRs. This paper reviews the graphite crystalline structure, the fabrication procedures, characteristics, chemical kinetics and modes of oxidation of nuclear graphite for future model developments.
... Table III summarizes the graphite oxidation models used in this calculation. These oxidation models are well validated by comparisons with the experimental data implemented into GAMMA code in the previous research (Lim and NO [13]). ...
The GAMMA+ code was originally developed for the system transient and safety analysis of a high temperature gas-cooled reactor (HTGR). Researches on the extension and improvement of the capability of the GAMMA+ code for applications to a sodium-cooled fast reactor (SFR) have been progressive. This paper summarizes the major items achieved on improvements of the GAMMA+ code to simulate thermo-fluid phenomena in an SFR. These include thermo-physical properties of sodium, empirical models and correlations (i.e., pressure drop and heat transfer correlations), a simplified model for stratified sodium and gas flow, and reactivity feedback models dedicated to an SFR. Such items for the improvement are essential to accommodate the features of an SFR which are different from those of an HTGR. The verification and validation of the improved version of GAMMA+ have been in progress. Initial efforts to verify and validate the improved version of GAMMA+ are also summarized in this paper. The topics of the efforts include conceptual problems having analytical solutions, heat exchanger performances, and a plant trip test of Monju. The results of the verification and validation studies demonstrate the good performance and reliability of the improved version of the GAMMA+ code for analyzing one-dimensional (1-D) as well as multi-dimensional thermo-fluid phenomena in an SFR.
Korea Atomic Energy Research Institute (KAERI) and Argonne National Laboratory (ANL) conducted steady-state tests at constant heat flux conditions to investigate the scale effect on heat removal behavior of the air-cooled Reactor Cavity Cooling System (RCCS). Two differently scaled-down test facilities were used in this study, KAERI’s ¼ scale with 4.0-m cavity height, and ANL’s ½ scale with 6.8-m cavity height. Two scaling laws were proposed to simulate the radiation across the cavity and the buoyancy-driven duct flow in the riser, respectively. The test matrix for KAERI and ANL test facilities focused on the scaling effect of the heated riser length with constant heat flux at the PMR200 RCCS design. The test results showed that mixed convection in the riser duct is an important factor for accurately extrapolating the thermo-fluid behavior in the prototype from the test results in the scale-down facilities. The system analysis code, GAMMA+, with improved heat transfer models predicted fairly well the air-cooled RCCS test data from two facilities. GAMMA+ analysis results showed that the predicted radiation fractions on the heated plate in the scale-down test conditions were larger than those on the reactor vessel in the prototype. The scaling law for air-cooled RCCS was improved by considering the mixed convection in the riser duct and the same radiation fraction on the heated plate. The mixed convection effect was calculated by the height ratio. The same radiation fraction provided a conservative extrapolation from the scale-down test results.
The four different geometries of a reactor vessel auxiliary cooling system (RVACS) are designed to investigate how the heat removal capability changes by the geometries. Each geometry has different heat transfer characteristics regarding airflow and the presence of an air separator and an insulation material. The heat removal performance is evaluated with the reactor vessel (RV) temperature, the RV wall emissivity, and the airflow gap. For the same RVACS volume, 600 °C RV temperature, and 6-cm gap, the heat removal capability varies from 240.8 kW to 308.5 kW, depending on the geometry. The wall emissivity is less effective for Geometry 2, which has a large cavity volume and a small heat transfer area compared to the other geometries. The highest heat removal performance was obtained using Geometry 3 because cold air flows in from the bottom of the RVACS, improving both radiative and convective heat transfer. Reducing the gap size by 3 cm results in only 80.0 kW of heat removal capability for Geometry 1, and the heat removal dramatically decreases to 0.2 kW at an RV temperature of 800 °C. A sufficient RVACS gap size of at least 6 cm with a 0.6-m diameter intake pipe is required to provide adequate natural circulation and eventually enhance the heat removal capability.
A helium circulator and a Printed Circuit Heat Exchanger (PCHE) type recuperator were developed for the helium cooling system (HCS) of a helium cooled breeding blanket for fusion reactor. Their full-scale components were manufactured and installed in the HeSS (Helium Supply System) facility at KAERI to perform qualification and performance tests. These components were tested to obtain performance curves for the circulator, and the effectiveness factor of the recuperator. Their component models were developed in GAMMA-FR based on design parameters and specifications. The results of performance testing were compared with corresponding simulations using GAMMA-FR to verify the components models. The simulation results shows good agreement with the test data. However, performance tests with a wide range of test conditions are required to better understand the performance of the components and further verification and validation of the models will be followed.
Supercritical carbon dioxide (sCO2) Brayton cycle-based power plants are being extensively explored as viable alternatives to traditional Rankine based steam power plants. Higher temperatures at the turbine exit in a sCO2 cycle provide better opportunities for heat recuperation, thus improving overall cycle efficiency. Among the various heat exchanger configurations available, compact microchannel heat exchangers commonly known as Printed Circuit Heat Exchangers (PCHE's) are typically used in sCO2 power plants. In the current work, a hybrid approach comprising of a Thermal Resistance Network (TRN) model coupled with a CFD model for estimating local heat transfer and pressure drops is presented for a PCHE core with straight and zigzag channel configurations. Full-scale TRN model is used to optimize the overall stack dimensions based on minimum rate of heat loss from the external surfaces of the PCHE core. The TRN model accounts for the thermo-physical property variations of sCO2 along the channel length to effectively capture the channel pressure drop and heat transfer. This is achieved by discretizing the heat transfer domain comprising of alternatively stacked hot and cold streams into sub-heat exchangers to account for variations in thermophysical properties while calculating the nodal friction factors and local heat transfer coefficients. CFD simulations are performed for a full length of a single stack of hot and cold fluid streams to arrive at corrected heat transfer and pressure drop correlations. Thermo-hydraulic analysis is performed for a range of channel hydraulic diameters and channel mass flow rates for both straight and zig-zag configurations to deduce optimum stack dimensions. The efficacy of the hybrid model is demonstrated with a case study of a counterflow recuperator used in a simple recuperated 1 MW sCO2 Brayton power plant.
Due to small footprint and high efficiency, a Supercritical CO2 (S-CO2) power cycle is considered to be one of the promising next generation power cycles. Although S-CO2 is reported as a powerful cleaning agent, the performance of a power system operating with S-CO2 will also inevitably degrade over time. Previous researchers have shown that turbomachinery deterioration could be a major subject regarding the system performance degradation. Nevertheless, no quantitative evaluation has yet been made so far. In this study, the impact of the performance degradation of the S-CO2 turbomachinery on the overall performance of the system is analyzed quantitatively. The concept of Health Parameter is used to simulate turbomachinery degradation. In order to quantify the impact, an S-CO2 direct-cycle small modular reactor is selected as a target system. A transient analysis platform is built using a nuclear system safety analysis code and the Deep Neural Network (DNN) based S-CO2 turbomachinery off-design performance model. System dynamics are evaluated for primary frequency control ability and secondary load-following capability. Results shows that the control problems during the transient state can occur when the output fluctuations are large and the performance degradation is severe. It has been confirmed that even control failures of the PID controllers can occur. Therefore, performance degradation of turbomachinery must be monitored and considered, for an operation strategy for S-CO2 systems.
The very high-temperature gas-cooled reactor (VHTR) has attracted interest owing to its high passive safety. A reactor analysis tool is required to analyze the safety characteristics of a VHTR in detail, and such a tool should be capable of a transient calculation for a reactivity insertion accident such as control rod ejection or withdrawal. This work presents a transient analysis of a block-type VHTR core. Time-dependent neutron diffusion equation is solved by the finite element method. A simplified thermal fluid analysis tool is also implemented to consider thermal feedback. In addition, a new method is introduced to resolve the control rod cusping effect without additional computation or mesh reconstruction. The above methods are applied to a reactor physics code CAPP, developed at Korea Atomic Energy Research Institute (KAERI). The methods are tested on three-dimensional VHTR problems that include control rods and show encouraging results.
The core of a prismatic very high temperature gas-cooled reactor (VHTR) is composed of stacked graphite blocks with gaps between them, which results in undesired flows through the gaps. These flows complicate the flow distribution in the reactor core and cause difficulty in predicting the temperature distribution of the graphite block. Conventionally, computational fluid dynamics (CFD) codes have been mainly used for the VHTR reactor core analysis. However, they require considerable calculation time and cost, and, therefore, are considered too expensive in terms of calculation time to investigate the effect of the gap size distribution in the core. As numerous cases with different gap size combinations need to be tested in reactor design, it can be said that high calculation speed of the design code with reasonable accuracy is an important feature.
In this study, a thermo-fluid analysis code for the core of a prismatic VHTR, named FastNet (Flow Analysis for Steady-state Network), was developed for prediction of the core flow and temperature distribution with affordable computational cost. For rapid calculation, a flow network analysis method was used for flow distribution analysis, and a thermal analysis model was added to analyze the whole core temperature distribution. To overcome the drawbacks of its low resolution, an effective thermal conductivity model and a maximum fuel temperature model were applied. Finally, to verify the code, results of the FastNet calculation were compared to other codes such as the CFD code and CORONA code as a code-to-code validation. The results show that a satisfactory accuracy was obtained with a remarkably short computational time.
A significant challenge in the core modeling of pebble bed reactors (PBRs) is the complex fuel-coolant structure. At the expense of approximating local flow and heat transfer effects, porous media models can provide medium-fidelity predictions of complicated thermal-fluid systems with significantly less computational cost than high-fidelity Computational Fluid Dynamics (CFD) models. This paper presents a new porous media code, Pronghorn – a fast-running core simulator intended to accelerate the design and analysis cycle for PBRs and provide boundary conditions for systems-level analysis. This paper describes the physical models in Pronghorn and demonstrates the capability of a friction-dominated model for predicting gas-cooled PBR decay heat removal by presenting simulation results for all 52 of the steady-state axisymmetric German SANA experiments, which include two different fluids and three different types of pebbles. The pebble temperature in all 52 cases is predicted with a mean error (predicted minus experimental) of +22.6 °C with standard deviation of 54.6 °C. To demonstrate Pronghorn's capability for modeling bed-to-plenum heat and mass transfer, one open-plenum SANA experimental case is also simulated. A code-to-code comparison with Flownex and GAMMA shows that Pronghorn is comparable in accuracy to other porous media simulation tools, with the additional advantages of 1) an arbitrary equation of state; 2) 3-D unstructured mesh capabilities; and 3) multiphysics coupling to other Multiphysics Object-Oriented Simulation Environment (MOOSE) applications. Finally, the effect of several porous media closure selections, in particular the porosity, the near-wall treatment for effective solid thermal conductivity, the interphase drag and heat transfer, and the fluid thermal dispersion, on temperature predictions are quantified.
A hydrogen production system coupled to very high temperature gas-cooled nuclear reactor (VHTR) is considered to be one of the most promising ways to achieve massive hydrogen production. The core of the prismatic VHTR consists of hexagonal graphite fuel blocks and reflector blocks. There exist gaps between graphite blocks; the vertical gap is referred to as a bypass gap and the horizontal gap is referred to as a cross gap. The coolant flows through these gaps as well as the coolant channels, thereby forming a complicated flow field in the core.
A looped network analysis method has been introduced for fast and simple flow distribution analysis. It has great strengths in analyzing a complex network in a short computational time; however, the looped network analysis has been used generally to analyze two-dimensional flow networks in practical applications. Because the flow network of the core of prismatic VHTR is three-dimensional, a new methodology to apply the looped network analysis to a three-dimensional flow network was proposed in this study.
This paper introduces the geometrical configuration of the reactor core of the prismatic VHTR, the looped network analysis method and its extension to the three-dimensional flow network, applied constitutive relations for closure and finally, the validation results against available multi-block bypass flow experimental data.
Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be operated at elevated temperature with high pressure helium coolant during normal operation in ITER. One of the main ancillary systems of HCCR-TBS is Helium Cooling System (HCS) which play an important role to extract heat from HCCR Test Blanket Module (TBM) by the helium coolant to keep the operational temperature and the extracted heat is finally transferred to ITER CCWS-1 (Component Cooling Water System) by a Printed Circuit Heat Exchanger (PCHE) in the HCS. In such circumstances if Loss Of Coolant Accident (LOCA) occurs in the PCHE, the high pressure helium coolant in the primary side goes into the lower pressure water in the secondary side thus pressurizing CCWS-1. In addition, since the helium coolant contains tritium due to permeation from the TBM, tritium migrates into CCWS-1, a non-nuclear system. In this paper, accident analysis for LOCA in the heat exchanger is presented. For the analysis, GAMMA-FR code which has been developed for fusion applications was used. Main components in the HCS and CCWS-1 were modelled as volume and junctions. The accident analysis was performed for the reference case with ten channels rupture and sensitivity study was also performed by changing the crack size. The results show that pressure and tritium requirement of CCWS-1 can be met in spite of LOCA in the heat exchanger of the HCCR-TBS HCS.
A hydrogen production system coupled to High Temperature Gas-cooled nuclear Reactor (HTGR) is considered to be one of the most promising ways for massive hydrogen production. For the reliability of the coupled system, the safety analysis on the HTGR is to be conducted by a system-scale analysis code. The system-scale analysis code adopts an effective thermal conductivity (ETC) model for a fuel block due to its complex geometry containing large number of coolant holes and nuclear fuel rods. The ETC of the fuel block is crucial to calculate the heat transfer inside the reactor core and prediction of thermal distribution over the reactor core is the most significant for the safety analysis of HTGR. Therefore, the verification of the ETC model that contributes to the prediction is essential. This ETC model based on Maxwell's theory shows an inaccurate prediction when the configuration of the composite materials is not homogeneous. Since the geometry of Reserve Shutdown Control (RSC) fuel block of HTGR is not homogeneous due to a large RSC hole, the ETC model for RSC fuel block should be developed to improve the accuracy and reliability of the reactor system analysis code. In this study, the two ETC models for the RSC fuel block have been developed by the thermal network modeling. Computational fluid dynamic simulations with a real geometry were performed to evaluate the accuracy of the ETC models for the RSC fuel block. The comparative result between CFD analysis and the ETC model shows that the newly developed model predicts the effective thermal conductivity of RSC fuel block more accurately than the previous model.
There exists a large body of publications on neutronics or thermal-fluids analyses of a prismatic gas-cooled reactor core. They mainly focus on stand-alone analyses, although the neutronics and thermal-fluids behavior affect each other. In order to consider the interaction between the neutronics and thermal-fluids behavior, a coupled analysis is desirable. However, only a few studies are available in the open literature on the coupled analysis of a prismatic gas-cooled reactor core. In this work, a code system for a coupled neutronics and thermo-fluids simulation was developed using CAPP (reactor physics code) and GAMMA+ (thermo-fluid system code). Using the code system developed, coupled neutronics and thermo-fluid simulations were carried out for a prismatic very high temperature reactor (VHTR) core to extend the knowledge about the coupled behavior. In particular, the effect of the bypass flow modeling on the coupled behavior was analyzed.
Despite the growing interest in the supercritical CO2 (S-CO2) Brayton cycle, research on the cycle transient behavior, especially in case of CO2 compressor inlet condition variation near the critical point, is still in its early stage. Controlling CO2 compressor operation near the critical point is one of the most important issues to operate a S-CO2 Brayton cycle with a high efficiency. This is because the compressor should operate near the critical point to reduce the compression work. Therefore, CO2 compressor operation and performance data from the S-CO2 compressor test facility called SCO2PE (Supercritical CO2 Pressurizing Experiment) were accumulated. The data are obtained under various compressor inlet conditions. Furthermore, in this study, the validation of the gas system transient analysis code GAMMA was carried out by utilizing the experimental data of SCO2PE. To simulate the data by the GAMMA code, the code was revised to model the compressor performance. A transient case for reduction in cooling event was simulated with the facility and the experimental data were compared to the revised GAMMA code. The revised GAMMA code showed a reasonable performance and demonstrated the potential of the code for being used in a larger scale S-CO2 power system.
In Korea, the Very High Temperature Gas-Cooled Reactor (VHTR) PMR200 is being developed in the Nuclear Hydrogen Development and Demonstration project. Its core consists of hexagonal prism-shaped graphite blocks for the fuel and reflector, and each hexagonal fuel block contains 108 cylindrical coolant holes and 210 fuel compacts. Because of these holes and fuels, the heat transfer in lateral directions in the fuel blocks becomes very complicated. Especially in accident situations when forced convection is lost, the majority of the afterheat flows in the radial direction by conduction across the large number of coolant holes. Moreover, radiation heat transfer is supposed to be added to the radial heat transfer modes owing to the high temperature of the VHTR core. Because of these complexities in radial heat transfer, reliable modeling for effective thermal conductivity (ETC) is required in order to analyze the reactor core thermal behavior using lumped-parameter codes, which are often used to evaluate the integrity of nuclear fuel embedded in the graphite block. In this study, the ETC model adopted in the GAMMA+ code was introduced, and the adequacy of the model was assessed by the commercial computational fluid dynamics (CFD) code CFX-13. The results of the CFD analysis were consistent with the ETC model in general even if a slight disagreement was shown for the case of high temperature. From these analyses, it could be concluded that the ETC model adopted in the GAMMA+ code is an adequate model for the analysis of the PMR200 reactor core. Moreover, it was found that the effect of fuel gap can cause an overprediction of the ETC if the fuel compact thermal conductivity is larger than the applicable range of the model.
Through consideration of the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He-Cooled Molten Lithium (HCML) Test Blanket Module (TBM) for testing in the International Thermonuclear Experimental Reactor (ITER). To validate the safety of the HCML TBM design concept and guarantee high efficiency of the power conversion system, an evaluation of the heat transfer capability of the gas coolant in a high Reynolds number regime should precede this test. In this study, a thermal hydraulic test with a high-pressure nitrogen gas loop was performed and a thermal hydraulic analysis was carried out with the commercial CFD code Fluent 6.3.26 and the system code GAMMA (Gas Multicomponent Mixture Analysis) under the same test conditions. In the experiment, a single TBM First Wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, was used, and the test was performed at pressures of 11, 19 and 29 bar and under various flow rates ranging from 0.63 to 2.44 kg/min. As one-side of the mock-up was heated by a furnace heater at a constant temperature, the wall temperatures were measured by installed thermocouples, with the measured temperatures showing strong parity with code results simulated under the same test conditions. Even with the system code using the modified Dittus-Boelter correlation, which was developed under a different heating condition, the three-dimensional approach of the system code is capable of estimating a one-sided heating condition in a fusion application.
The Korea nuclear industry has been developing the thermal-hydraulic system analysis Safety and Performance Analysis CodE (SPACE) and the GAs Multicomponent Mixture Analysis (GAMMA) code for safety analysis of pressurized water reactors (PWRs) and hightemperature gas-cooled reactors (HTGRs), respectively. SPACE will replace outdated vendor-supplied codes and will be used for the safety analysis of operating PWRs and for the design of an advanced PWR. SPACE consists of up-to-date physical models of two-phase flow dealing with multidimensional two-fluid, three-field flow. GAMMA consists of multidimensional governing equations consisting of the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of n species. GAMMA is based on a porous media model so that the rmofluid and chemical reaction behaviors in a multicomponent mixture system and heat transfer within solid components, free and forced convection between a solid and a fluid, and radiative heat transfer between solid surfaces can be dealt with. GAMMA has a two-dimensional helium turbine model based on the throughflow calculation and a coupled neutronics-thermal-hydraulic model. Extensive code assessment has been performed for the verification and validation of SPACE and GAMMA.
In Korea, the Very High Temperature Gas-Cooled Reactor (VHTR) PMR200 is being developed in the Nuclear Hydrogen Development and Demonstration project. Its core consists of hexagonal prism-shaped graphite blocks for the fuel and reflector, and each hexagonal fuel block contains 108 cylindrical coolant holes and 210 fuel compacts. Because of these holes and fuels, the heat transfer in lateral directions in the fuel blocks becomes very complicated. Especially in accident situations when forced convection is lost, the majority of the afterheat flows in the radial direction by conduction across the large number of coolant holes. Moreover, radiation heat transfer is supposed to be added to the radial heat transfer modes owing to the high temperature of the VHTR core. Because of these complexities in radial heat transfer, reliable modeling for effective thermal conductivity (ETC) is required in order to analyze the reactor core thermal behavior using lumped-parameter codes, which are often used to evaluate the integrity of nuclear fuel embedded in the graphite block. In this study, the ETC model adopted in the GAMMA+ code was introduced, and the adequacy of the model was assessed by the commercial computational fluid dynamics (CFD) code CFX-13. The results of the CFD analysis were consistent with the ETC model in general even if a slight disagreement was shown for the case of high temperature. From these analyses, it could be concluded that the ETC model adopted in the GAMMA+ code is an adequate model for the analysis of the PMR200 reactor core. Moreover, it was found that the effect of fuel gap can cause an overprediction of the ETC if the fuel compact thermal conductivity is larger than the applicable range of the model.
The GAMMA+ and Flownex codes are both based on a one-dimensional flow network modelling approach and both can account for any complex network of different heat transfer phenomena occurring simultaneously. However, there are notable differences in some of the detail modelling aspects, such as the way in which the convection in the reactor cavity is represented. Despite this, it was found in the analyses of the air-cooled RCCS system that the results provided by the two codes compare very well if similar input values are used for the pressure drop coefficients, heat transfer coefficients and view factors.
The results show that the radiation heat transfer comprises the bulk of the total rate of heat transfer from the RPV surface. It is also shown that it is possible to obtain a stable and sustainable steady-state operational condition where the flow is in the reverse direction through the RCCS standpipes, resulting in excessively high values for the concrete wall temperature. It is therefore crucial in the design to ensure that such a flow reversal will not occur under any circumstances.
In general the good comparison between the two codes provides confidence in the ability of both to correctly solve the fundamental conservation and heat transfer relations in an integrated manner for the complete RCCS system. Provided that appropriate input values are available, these codes can therefore be used effectively to evaluate the integrated performance of the system under various operating conditions. It is shown here that the RCCS should remain functional and continue to provide sufficient cooling even for very high blockage ratios at the inlet to the riser ducts, which supports the safety case.
Korea has designed a helium cooled ceramic reflector (HCCR) based test blanket system (TBS) for an ITER. An in-vessel loss of coolant accident is one eight selected reference accidents in the Korean TBS. This accident is initiated by a single or multiple rupture of the test blanket module first wall cooling channels, causing a plasma disruption, and pressurization of the vacuum vessel (VV). In this type of accident, the governing parameters are various, for example, the operating pressure, gas temperature, TBS volume, VV volume, and mass flow rate. Thus, a scoping study is an essential strategy when attempting to determine the proper design specification for a Korean TBS. In this paper, given the preliminary accident analysis results for the current HCCR TBS, a parametric study was performed. For this transient simulation, the Korean nuclear fusion reactor safety analysis code (GAMMA-FR) was used.
The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V&V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of well-defined multi-dimensional computational benchmark problems with a common given set of cross-sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants' results are compared using a statistical method and possible areas of future code improvement are identified.
We present three nuclear/hydrogen-related R&D activities being performed at KAIST: air-ingressed LOCA analysis code development, gas turbine analysis tool development, and hydrogen-production system analysis model development. The ICE numerical technique widely used for the safety analysis of water-reactors is successfully implemented into GAMMA, with which we solve the basic equations for continuity, momentum conservation, energy conservation of the gas mixture, and mass conservation of 6 species (He, N2, O2, CO, CO2, and H2O). GAMMA has been extensively validated using data from 14 test facilities. We developed a tool to predict the characteristics of HTGR helium turbines based on the throughflow calculation with a Newton-Raphson method that overcomes the weakness of the conventional method based on the successive iteration scheme. It is found that the current method reaches stable and quick convergence even under the off-normal condition with the same degree of accuracy. The dynamic equations for the distillation column of HI process are described with 4 material components involved in the HI process: H2O, HI, I2, H2. For the HI process we improved the Neumann model based on the NRTL (Non-Random Two-Liquid) model. The improved Neumann model predicted a total pressure with 8.6% maximum relative deviation from the data and 2.5% mean relative deviation, and liquid-liquid-separation with 9.52% maximum relative deviation from the data.
KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.
Korea has designed a Helium-Cooled Ceramic Reflector (HCCR)-based Test Blanket System (TBS) for International Thermonuclear Experimental Reactor (ITER). Among seven selected reference accidents in Korean TBS, in-box loss of coolant accident (LOCA) is one of them. This is initiated by a double-ended break of the coolant pipe in the Breeding Zone (BZ), pressurizing the BZ box structure, causing pressurization of the Tritium Extraction System (TES) and purging of pipelines. When the accident is detected, the Plant Safety System (PSS) isolates the Helium Cooling System (HCS) and TES, and requests plasma shutdown to Fusion Power Shutdown System (FPSS). To prevent aggravating failure of the system, the safety function is automatically activated when the accident is detected, the device being the isolation valve of HCS and TES. One important observation of this accident is that instant isolation is not a good measure to take. In terms of the possibility of aggravating failure, system isolation is an important safety procedure but isolated TES volume is exposed to high pressure and temperature conditions in the early move of the accident transient. The result of system safety analysis shows that delayed isolation keeps the system safe for a while. In this article, given the preliminary accident analysis results for the current HCCR TBS, case studies were performed regarding the delayed isolation timing effect. For this transient simulation, Korean nuclear fusion reactor safety analysis code (GAMMA-FR) was used.
Massive ingress of air into the core of a high-temperature gas cooled reactor is among the accidents with a low occurrence frequency, but there are still gaps in understanding with respect to its consequences. In the present paper, massive air ingress combined with a delayed start of the afterheat removal system is investigated and compared to air ingress accidents with normal operation of the afterheat removal procedure. A computer programme REACT/THERMIX used for these accident analyses is described. For a high-temperature gas cooled reactor with a pebble bed core, it is shown that massive air ingress has no real safety endangering consequences even if the operation of the afterheat removal system is delayed by 6 h.
Experiments and numerical analyses on mass transfer in a gas mixture laminar flow through a circular graphite tube accompanied with graphite oxidation were performed to clarify characteristics of graphite corrosion in connection with the primary cooling pipe rupture accident in a high temperature gas cooled reactor. In the experiment, inlet Reynolds numbers ranged from 16 to 320, graphite temperatures from 600 to 1,050°C and inlet oxygen mass fractions from 10 to 50%. IG-110 and PGX graphites were used. The mass transfer coefficients in the experiment at a high temperature were less than those obtained on the basis of analogy between heat and mass transfer. It was found from the numerical analysis that the reduction of mass transfer coefficient resulted from the injection flow from the wall. The mass fractions of carbon monoxide and carbon dioxide in the experiment were well predicted on the basis of the one-dimensional numerical analysis in which graphite oxidation and carbon monoxide combustion were taken into consideration.
The ICE technique for numerical fluid dynamics has been revised considerably, and generalized in such a way as to extend the applicability to fluid flows with arbitrary equation of state and the full viscous stress tensor. The method is useful for the numerical solution of time-dependent fluid flow problems in several space dimensions, for all Mach numbers from zero (incompressible limit) to infinity (hypersonic limit). This new version is considerably less complicated than the original form. The present description does not assume a familiarity with the previous one.
This paper reports on the TINTE code. It deals with the nuclear and thermal transient behavior of a high-temperature reactor, taking into consideration the mutual feedback in two-dimensional r-z geometry including the following: time-dependent neutron flux calcultion; time-dependent heat source distribution (local and nonlocal fractions); time-dependent heat transport from the fuel to the fuel element surface; time-dependent global temperature distribution; gas glow distribution for a given mass flow or given mass flow or given pressure difference or even under natural circulation conditions; and convesction and its feedback to the circulation. The TINTE code has been tested against transient experiments. It has also been used successfully for other reactor investigations and theremofluid problems including design studies for a small heating reactor.
Measurements of the diffusion coefficient for a trace of carbon dioxide in a mixture of helium and nitrogen of varying composition have been made at room temperature and atmospheric pressure using the point source technique. The results are shown to be in agreement with the results of Fairbanks and Wilke in that an equation of the form Di−mix=[∑ki≠k(xk/Dik)]−1 can be used to relate the diffusion coefficient of a trace of species i in a mixture of composition given by the mole fraction xk to the binary diffusion coefficients Dik. Some data are presented on the separation of the mixture in the diffusion zone.
Massive ingress of air into the core of a high-temperature gas cooled reactor is among the accidents with a low occurrence frequency, but there are still gaps in understanding with respect to its consequences. In the present paper, massive air ingress combined with a delayed start of the afterheat removal system is investigated and compared to air ingress, accidents with normal operation of the afterheat removal procedure. A computer programme REACT/THERMIX used for these accident analyses is described. For a high-temperature gas cooled reactor with a pebble bed core, it is shown that massive air ingress has no real safety endangering consequences even if the operation of the afterheat removal system is delayed by 6 h.
The oxidation of moist carbon monoxide and the post-induction-phase oxidation of methane were studied in a turbulent flow reactor. Reactants, stable intermediates, and products were determined spatially by chemical sampling and gas-chromatographic analysis.The carbon monoxide-oxygen reaction in the presence of water was studied at atmosphericpressure, and over the following ranges: temperature, 1030°–1230°K; equivalence ratio, 0.04–0.5; and water concentration, 0.1%–3.0%. The over-all rate expression found was −d[CO]/dt=1014.6±0.25 exp[(−40,000±1200)/RT][CO]1.0[H2O]0.5[O2]0.25 mole cm−3 sec−1. The data support the fact that hydroxyl radical concentration in the reaction exceeds that at thermal equilibrium by as much as 2 orders of magnitude.The post-induction-phase reaction of methane and oxygen was studied at atmospheric pressure, over the temperature range of 1100°–1400°K and equivalence ratio range of 0.05–0.5. The over-all methane disappearance-rate expression was found to be −d[CH4]/dt=1013.2±0.20 exp[(−48,400±1200)/RT][CH4]0.7[O2]0.8, mole cm−3 sec−1. The rate was shown to be independent of water concentrations added initially or produced in the reaction.The over-all appearance rate of carbon dioxide in the methane-oxygen reaction is described byd[CO2]/dt=1014.75±0.40 exp[(−43,000±2200)/RT][CO]1.0[H2O]0.5[O2]0.25 mole cm−3 sec−1. This correlation represents rates of carbon dioxide formation 3.5 times slower than those found in the independent study of the moist carbon monoxide reaction.From these and other experiments it was possible to deduce thatCH4+OH→CH3+H2O (3) is not the only mechanism contributing to the observed rate of disappearance of methane. It was concluded that the reaction CH4+O→CH3+OH (4) is of major importance in both oxygen-and fuel-rich systems at high temperatures. Furthermore, the experimental data support that these two reactions, as well as CH3+H2→CH4+H (−5) contribute to the methane results reported here.
Air Ingress Behavior during a Primary-Pipe Rupture Accident of HTGR
- T Takeda
Mixing Process of a Binary Gas in a Density Stratified Layer
- T Takeda
DIREKT-AComputer Programme for Non-Steady, Two-Dimensional Simulation of Thermo-Hydraulic Transients
- S Struth
Multi-Dimensional Multicomponent Mixture Air-Ingress Analysis and Air-Ingress Experiment in an HTGR
- H C No
- E S Kim
- H S Lim
The Finite Element Version of TRIO Code
- J P Magnaud
- S Goldstein