Thermal-hydraulic System Analysis Code Research Articles

CFD Applications to Pressurized Thermal Shock-Related Phenomena

In pressurized water reactor accident scenarios, the injection of water from the emergency core cooling system (ECCS) (ECC injection) might induce a pressurized thermal shock (PTS), affecting the reactor pressure vessel (RPV) integrity. Therefore, PTS is a vital research issue in reactor safety, and its analysis is essential for evaluating the integrity of RPVs, which determines the reactor life. The PTS analysis comprises a coupled analysis between thermal–hydraulic and structural analyses. The thermal–hydraulic approach is particularly crucial, and reliable computational fluid dynamic (CFD) simulations should play a vital role in the future because predicting the temperature gradient of the RPV wall requires data on the transient temperature distribution of the downcomer (DC). Since one-dimensional codes cannot predict the complex three-dimensional flow features during ECC injection, PTS is one reactor safety issue where CFD simulation can benefit from complement evaluations with thermal–hydraulic system analysis codes. This study reviewed from the viewpoint of the turbulence models most affecting PTS analysis based on papers published since 2010 on single- and two-phase flow CFD simulation for the experiment on PTS performed in the Rossendorf coolant mixing model (ROCOM), transient two-phase flow (TOPFLOW), upper plenum test facility (UPTF), and large-scale test facility (LSTF). The results revealed that in single-phase flow CFD simulation, where knowledge and experience are sufficient, various turbulence models have been considered, and many analyses using large eddy simulation (LES) have been reported. For two-phase flow analysis of air–water conditions, interface capturing/tracking methods were used in addition to two-fluid models. The standard k–ε and shear stress transport (SST) k–ω models were still in the validated phase, and various turbulence models have yet to be fully validated. In the two-phase flow analysis of steam–water conditions, many studies have used two-fluid models and Reynolds-averaged Navier-Stoke (RANS), and NEPTUNE_CFD, in particular, has been reported to show excellent prediction performance based on years of accumulated validation.

Development of thermo-physical property database and code of fusion materials for CFETR blankets

The conceptual design of China Fusion Engineering Test Reactor (CFETR) has been completed and is currently in the engineering design phase. In this paper, focusing on three categories of candidate blanket designs and divertor designs, the selection schemes for materials of plasma-facing components, solid structure, tritium breeding materials, neutron multiplier, coolant, and divertor were extracted. Thermo-physical property data was compiled to form a material thermo-physical property database, named Thermo-physical Property Database (ThePD), which is the first advanced thermo-physical database for materials of CFETR blankets. Besides, the callable code was generated, which can be directly applied in the system code for the blanket. Based on internationally available database from National Institute of Standards and Technology (NIST) and similar software AP1700, the correctness of the data and the transitional analysis of regions designed by International Association for the Properties of Water and Steam (IAPWS) were verified. Using the thermal-hydraulic system analysis code, integrated validation was performed for the validated thermo-physical property data and code, demonstrating the applicability and accuracy of the data. ThePD is conducive to addressing the issue of incomplete thermo-physical property data faced by blanket design and safety analysis software for CFETR, and it can fill this gap.

Analysis of control rod driving mechanism nozzle rupture with loss of safety injection at the ATLAS experimental facility using MARS-KS and TRACE

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

Analysis of 4-inch cold leg top-slot break LOCA in ATLAS experimental facility using MARS-KS

Abstract Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) with reference to the Advanced Power Reactor 1400 MW (APR1400) for transient and design basis accidents (DBAs) simulation. An experiment for a 4-inch cold leg top-slot break was performed at ATLAS to resolve a safety issue that the loop seal reformation (LSR) of APR1400 could lead to the increase of peak cladding temperature. In addition, the experimental data has been utilized to validate system codes within a framework of domestic standard problem (DSP) program organized by KAERI in collaboration with Korea Institute of Nuclear Safety (KINS). In this study, the experiment has been analyzed by thermal-hydraulic system analysis code, MARS-KS 1.5 and a comparison with experimental and calculation results has been performed. Since the top-slot break is not a typical break geometry for safety analyses, this study aims at examining the applicability of MARS-KS to the top-slot break accident where the LSR occurs repeatedly. The results revealed that overall physical behavior during the accident was predicted by the code, appropriately. MARS-KS showed the excursion of the peak cladding temperature because of the LSR as in experiment. It has been confirmed that the core integrity was maintained because the temperature excursion by the LSR was not large enough to alter the acceptance criteria. In addition, it is presented the results of the sensitivity analysis of parameter that affect the figure of merits.

Numerical Evaluation of Coolability Limits of External Reactor Vessel Cooling Using an Improved Thermal-Hydraulic System Analysis Code

The external reactor vessel cooling provides a means to stabilize the molten core inside the reactor vessel during the severe accident in nuclear power plants. For a credible evaluation of the chances of success of in-vessel melt retention, one requires high-fidelity predictions on the natural circulation flow around the reactor vessel and, in particular, the corresponding critical heat flux (CHF) on a hemispherical lower head. In this study, the coolability limit of external reactor vessel cooling in the Korean APR1400 reactor was investigated numerically by improving a thermal-hydraulic system analysis code, MARS-KS1.5. Since such system analysis codes did not incorporate a CHF model for a downward-facing hemisphere, three CHF correlations were newly implemented in the MARS-KS code, two of them taking into account the effect of the mass flux. A transient analysis on the CHF and surface heat flux on the lower head was carried out while the surrounding cooling water was heated to create the two-phase natural circulation flow. The heat transfer mode on the heated surface was used as a measure of evaluating the success of the strategy, and the simulation results were compared with those obtained via the default CHF model in the code. The sensitivity analysis on the heat load from the core melt was also performed.

Comparative study of constitutive relations implemented in RELAP5 and TRACE – Part I: Methodology & wall friction

Nuclear thermal-hydraulic system analysis codes have been developed to simulate nuclear reactor systems, which solve simplified governing equations by replacing source terms with constitutive relations for simulating entire reactor systems with low computational resources. For half a century, many efforts have been made for wider versatility and higher accuracy of system codes, but various factors can affect the code analysis results, and it was difficult to isolate these factors and interpret them individually. In this study, two system codes, RELAP5 and TRACE, which have many users and are highly reliable, are selected to analyze only the effects of constitutive relations. The influence of constitutive relations is analyzed using in-house platforms that replicate constitute relations of RELAP5 and TRACE equally to exclude factors that may affect analysis results, such as governing equation solvers and user effects. Among the various constitutive relations, the analysis is performed on the wall variables expected to have the most influence on the analysis results. Part 1 paper presents the methodology and wall friction model comparison, while Part 2 paper shows wall heat transfer comparison of the two selected codes.

Development of a special thermal-hydraulic component model for the core makeup tank

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

Comparative study of constitutive relations implemented in RELAP5 and TRACE – Part II: Wall boiling heat transfer

Nuclear thermal-hydraulic system analysis codes have been developed to comprehensively model nuclear reactor systems to evaluate the safety of a nuclear reactor system. For analyzing complex systems with finite computational resources, system codes usually solve simplified fluid equations for coarsely discretized control volumes with one-dimensional assumptions and replace source terms in the governing equations with constitutive relations. Wall boiling heat transfer models are regarded as essential models in nuclear safety evaluation among many constitutive relations. The wall boiling heat transfer models of two widely used nuclear system codes, RELAP5 and TRACE, are analyzed in this study. It is first described how wall heat transfer models are composed in the two codes. By utilizing the same method described in Part 1 paper, heat fluxes from the two codes are compared under the same thermal-hydraulic conditions. The significant factors for the differences are identified as well as at which conditions the non-negligible difference occurs. Steady-state simulations with both codes are also conducted to confirm how the difference in wall heat transfer models impacts the simulation results.

Study on Calculation Method of Courant Limit in Thermal Hydraulic System Analysis Code

In order to analyze the influence of different Courant limit calculation methods on the calculation speed and accuracy of reactor thermal hydraulic system analysis code, two kinds of Courant limit calculation methods were studied: synthesis method and grouping method. The calculation principles of the two methods were analyzed, and the two methods were adopted to calculate the steady-state condition and large break loss of coolant accident condition of a PWR respectively. The results show that there is no significant difference between the two methods in steady-state condition; in the case of large break loss of coolant accident with drastic change of velocity field, the synthesis method can obtain more accurate calculation results, but it takes more time, and the grouping method can obtain faster calculation speed, however its calculation accuracy is lower.

Drift-flux parameter modeling of vertical downward gas–liquid two-phase flows for interfacial drag force formulation

The two-fluid model has been adopted as a platform of nuclear thermal–hydraulic system analysis code because it can treat the mechanical and thermal non-equilibrium between phases through the interfacial transfer terms. Precise modeling of the area-averaged interfacial drag force in the interfacial momentum transfer term is essential in predicting void fraction accurately. The drift-flux parameters, such as the distribution parameter and drift velocity, play an essential role in formulating the area-averaged interfacial drag force. This study aims at developing a drift-flux correlation with a wide application range for vertical downward two-phase flows. First, over 1200 experimental void fraction data of vertical downward two-phase flows were collected from 13 sources. Then, the existing correlations of the distribution parameter and drift velocity were reviewed and examined. Finally, a new drift-flux correlation was developed based on a large amount of experimental data and state-of-the-art knowledge of two-phase flow behaviors. The comparison between the experimental and calculated void fractions by the newly-developed correlation demonstrated that the new correlation could achieve superior performance to the existing correlations. More than 93% of the predicted void fractions were predicted within ± 20% error with the mean relative deviation and mean absolute relative deviation of 0.609% and 9.93%, respectively.

Effects of inclined condition on LOCA for a small offshore reactor with OTSG

A small offshore reactor called SMR-50 is conceptually designed for power supply in the ocean. The SMR-50 is a compact pressurized water reactor with two symmetrically arranged once-through-steam-generators (OTSGs). In this paper, the effect of inclined condition on the thermal–hydraulic behavior of the reactor during a postulated loss-of-coolant accident (LOCA) is analyzed. To simulate the transient under inclined condition, an analysis model is added into the thermal–hydraulic system analysis code RELAP/SCDAPSIM/MOD3.4. The simulation model of the reactor is developed and the LOCA transient under different inclined angles are simulated using the modified code. The results show that the inclined condition has a significant influence on thermal–hydraulic behavior of the reactor. Particularly, a rise of 500℃ in peak cladding temperate (PCT) is observed when inclined at a large angle in the plane of the two OTSGs. The inclined condition places a greater demand on safety systems of such a small offshore reactor and should be fully considered in the design process.

A numerical derivation of internal energy equation under a staggered mesh system for 1-D thermal hydraulic system analysis codes

This paper introduces a modified internal energy equation derived for multiphase flow in various flow conditions for a staggered mesh system. The pressure drop and heat dissipation terms in the internal energy conservation equation currently being developed based on the assumption that the scalar and momentum variables, which originate from the total energy and mechanical energy equations, respectively, are all located in the cell center of the control volume were redefined such that two different pressures and velocities stored in both cell centers and faces were imposed for the internal energy conservation. To achieve this, first, a modified internal energy conservation equation for a staggered mesh was derived by subtracting the mechanical energy equation from the total energy equation. The equation was then discretized classifying each term by its origin; variables that originated from the total energy equation were defined in the cell center, whereas terms that came from the mechanical energy equation were identified at faces. Since the discretized form of the proposed equation contained face velocity and cell pressure for the heat dissipation and pressure drop terms, respectively, these two terms were calculated implicitly leading to enhanced numerical stability. The accuracy of the modified internal energy equation in predicting the system pressure, fluid temperature, and heat balance for various flow channels was assessed. The verification of the proposed equation was completed through simulations of multiple theoretical problems including saturated liquid depressurization, adiabatic expansion of hydrogen, heat transfer in a helical steam generator, and energy transfer in a converging pipe. An improved result was obtained with the modified equation as the numerical calculation results agreed well with the analytic solutions with relative deviations less than 0.1% for most cases, while the solution obtained by the conventional internal energy equation showed a significant amount of deviation from the analytic solution.

Modeling and simulation of onset of condensation-induced water hammer

In view of the importance of the condensation-induced water hammer in various engineering fields, this study performed an analytical modeling and numerical simulation to obtain the map of the onset of the condensation-induced water hammer. Two linear boundaries were experimentally observed for the transitions between no condensation-induced water hammer and non-periodic condensation-induced water hammer and between non-periodic condensation-induced water hammer and periodic condensation-induced water hammer. The analytical modeling based on the energy balance equation was conducted to derive the onset of the condensation-induced water hammer. The analytical model suggested that the map of the onset of the condensation-induced water hammer could be displayed by two essential non-dimensional parameters, such as modified Jakob number and Froude number. The model indicated that the slope of the boundary of the onset of the condensation-induced water hammer was the product of modified Stanton number, non-dimensional interfacial area concentration, and non-dimensional temperature difference. The model could predict the boundaries between no condensation-induced water hammer and non-periodic condensation-induced water hammer and between non-periodic condensation-induced water hammer and periodic condensation-induced water hammer satisfactorily. The one-dimensional nuclear thermal-hydraulic system analysis code, RELAP5, was used for simulating the onset of the condensation-induced water hammer. The obtained boundary corresponded to the experimentally obtained boundary between non-periodic condensation-induced water hammer and periodic condensation-induced water hammer.

Counterpart test for LSTF 1% cold-leg break LOCA (SB-CL-32) utilizing ATLAS test facility

The counterpart test of the OECD-ATLAS project named A5.1 was performed using a thermal–hydraulic integral effect test facility, ATLAS. The target scenario for the A5.1 test was the LSTF SB-CL-32 test. The SB-CL-32 test is a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action. The test objective is to investigate the thermal hydraulic phenomena during a cold leg SBLOCA such as core heat-up, loop seal clearing (LSC), and the effect of AM actions. To achieve the test objectives, scaling analysis between LSTF and ATLAS was conducted first. The detailed test conditions such as initial and boundary conditions and sequence of events were determined based on the scaling analysis result. The experimental result showed that overall thermal hydraulic behavior which can be expected in this kind of SBLOCA scenario was observed. However, significant differences were found in the A5.1 test result compared with the LSTF SB-CL-32 test result in the view point of the loop seal clearing phenomenon and the maximum heater rod surface temperature behavior. These differences result from the different design between two test facilities which are refer to the different prototype nuclear power plant. Thus, several test condition parameters were analyzed as the sensitivity study utilizing the thermal hydraulic system analysis code to check their effect on the system response. From the sensitivity analysis result, it was confirmed that the A5.1 test was well modeled as a counterpart test of LSTF and realized the thermal hydraulic behavior during SBLOCA transient. This test result can be utilized to evaluate the prediction capability of thermal hydraulic system analysis codes and to validate the scale-up capability of integrated effect test (IET) test data to the nuclear power plant condition.

Integral Effect Test on Top-Slot Break Scenario With 4 Inches Cold Leg Break LOCA in ATLAS Facility

An experimental simulation for a loss of coolant accident with a 4 inch cold leg top-slot break was performed using ATLAS, an integral effect test facility at KAERI, Korea. The aims of this test are to resolve a safety issue and to validate the thermal hydraulic safety analysis codes for the activity of the 4th ATLAS-Domestic Standard Problem. The related safety issue for this test is that a loop seal clearing and loop seal reformation of APR1400 can lead to a core temperature excursion because of the characteristics of the loop seal geometry. In the experimental results, the core heater temperature slowly increased after loop seal reformation. This increase is attributed to the accumulated steam at the upper head. The accumulated steam increased the pressure and saturated the temperature in the core. The core heater temperature then increased due to the increased saturated temperature. This temperature increase therefore was not a core temperature excursion. Sensitivity analysis results using MARS-KS, which is a best-estimate thermal hydraulic system analysis code, are also presented in this study. In the 4th ATLAS-Domestic Standard Problem, 15 participants performed calculations and sensitivity analyses using various thermal hydraulic safety analysis codes (MARS-KS, SPACE, RELAP5, and TRACE). From this exercise, several parameters that affect the calculation results are brought out. The sensitivity analysis results using MARS-KS with some of these parameters are also presented. The MARS-KS calculation results are similar to the experimental data. The effects of critical flow model, break line modeling, and fine node modeling are also discussed.

Effect of interfacial drag force model on code prediction for upward adiabatic two-phase bubbly flow in vertical channels

Accurate modeling of the interfacial drag force is one of the keys to predicting thermo-fluid parameters using one-dimensional nuclear thermal-hydraulic system analysis code architected through the two-fluid model. The interfacial drag force appears in the interfacial momentum transfer term and governs the velocity slip or the relative velocity between gas and liquid phases. The most straightforward method to model the interfacial drag force is to model the force through the drag law (drag law approach). A drag coefficient and interfacial area concentration should be given to close the interfacial drag force model. Among them, the modeling of the interfacial area concentration has been one of the weakest links in the interfacial drag force modeling due to the lack of reliable data covering a wide test condition including prototypic nuclear reactor conditions and lack of physically sound interfacial area model. To avoid a considerable uncertainty in the prediction of the interfacial area concentration, Andersen and Chu (1982) proposed the interfacial drag force model using the drift-flux parameters (Andersen approach). The Andersen approach is practical for the simulation of a slow transient flow and a steady flow. Major system analysis codes such as USNRC TRACE have adopted the Andersen approach in the interfacial drag force modeling. Some attempts to improve the code performance have been considered using the drag law approach with the interfacial area transport equation. The dynamic modeling of the interfacial area concentration has the potential to improve the prediction accuracy of the interfacial area concentration in a transient flow and developing flow. Due to the importance of the improved interfacial drag force modeling, the implementation and evaluation of the interfacial area transport equation in USNRC TRACE code has been performed by Talley et al. (2011, 2013). The study claimed that the introduction of the interfacial area transport equation into the TRACE code improved the code performance in an adiabatic bubbly flow analysis significantly. The present study assessed the code calculation made by Talley et al. (2011) and identified several issues in the code calculation results. The present study analytically demonstrated that the drag law approach became identical with the Andersen approach for the distorted particle regime (or a major bubble shape regime in bubbly flow) due to the balancing-out of the interfacial area concentration (or bubble size) in the numerator and denominator of the interfacial drag force formulation. The code calculation using TRAC code endorsed the analytical assessment of the insignificant or no merit of the interfacial area transport equation in the code performance of the adiabatic bubbly flow analysis. The present study also pointed out the inconsistency of the code calculation made by Talley et al. (2011).

Simulation investigation of ocean motion on small reactor based on modified system code

Offshore nuclear small reactor is a promising research direction during recent years. In order to carry out system simulation and safety analysis under accident conditions for marine reactors, the classic thermal-hydraulic system analysis code RELAP5 is modified by taking heaving, inclining and rolling motion into consideration in this paper. Additional forces are added to the momentum equation for heaving and rolling, and control volume coordination are transformed for inclining and rolling. Then, a simplified small reactor with two loops is simulated by the modified code and the effect of ocean motion is investigated. Generally, heaving leads to the periodic fluctuation of mass flow rate, inclining would cause the decrease of total mass flow through the core, and the effect of rolling is more complex. Simulation analysis is also performed on two design basis accident LOCA and LOFA in this study. Safety related conclusions are obtained about the impacts of various ocean motion.

Research on code coupling method in severe accident simulator development

After Fukushima nuclear accident, the newly published nuclear regulation featured requests for the development of severe accident (SA) simulators in operator training. In response to such requests, this research deals with a code coupling method, the coupling of thermal-hydraulic system analysis code with integral SA code. Based on the existing full-scope simulator (FSS), the architecture of SA simulator was designed through the method. The crucial technical route is that the RELAP5/MOD3.2 code is utilized for early response, when the coupling condition is met, it switches to MAAP4.0.3 code to calculate middle-late accident. To ensure smooth transition, their major differences were analyzed and some coupling parameters were selected. A test case of SBLOCA was performed via RELAP5-couple-MAAP4, and some specific quantitative assessments were defined to justify the method. The results show a smooth transition between RELAP5 and MAAP4 is achieved and is expected to extend FSS function to SA field.

Application of adjoint based node optimization method to nuclear thermal-hydraulic system analysis code

Nuclear thermal-hydraulic system analysis codes with the Best Estimate plus Uncertainty approach are unavoidably uncertain and naturally, an uncertainty analysis has been an important issue. In general, the system is discretized depending on the user while satisfying some criteria. Previously, uncertainties related to spatial discretization in nodalization have been less considered due to their difficulty in quantification. However, there are some results that they are comparable to other uncertainties. Previous studies have been conducted by evaluating and optimizing the uncertainty related to user effect by performing iterative calculations. Since these methods consume a large amount of resources for computation, this study suggests a method of consuming less resources. For this purpose, the adjoint method is applied to the code for node optimization in this study. The proposed methodology is applied to two cases and confirmed its applicability for the node optimization and its utility in the node uncertainty quantification.

Node configuration uncertainty in nuclear safety analyses

In nuclear safety analyses, a 1-D thermal hydraulic system analysis code with the best estimate plus uncertainty (BEPU) approach is generally used. There are sources of uncertainty such as nodalization, physical models, and numerical scheme. Among them, the authors specifically focused on the node configuration uncertainty related to spatial discretization that is regarded as less dominant sources of uncertainty in the code previously. Furthermore, it was not easy to quantify the uncertainty from the node configuration. The authors quantified and analyzed the node configuration uncertainty to understand how much it affects the solution. In this study, the effect of spatial discretization are studied by discussing simulation results for various spatially discretized node systems. The discussion on the node configuration uncertainty will be presented by examples of SUbcooled BOiling (SUBO) test simulation and Loss Of Fluid Test (LOFT) simulation using Multi-dimensional Analysis of Reactor Safety (MARS) code developed in South Korea. From SUBO test simulation, it was found that when the number of nodes is small than the node configuration uncertainty can overwhelm the uncertainty from the physical model. From LOFT simulation, it was revealed that the node configuration can cause bifurcation in the complex nuclear system analyses. Thus, it is concluded that the node configuration uncertainty can be quite significant and has to be considered in the nuclear system analyses.