Thermal-hydraulic System Research Articles

Design and Control of a Fueled Molten Salt Cartridge Experiment for the Versatile Test Reactor

The mission of the Versatile Test Reactor (VTR) is to enable accelerated testing of advanced reactor fuels and materials as required for advanced reactor technologies. Each advanced reactor type has unique challenges, and these challenges affect the design of the testing vehicles used for accelerated testing. For molten salt reactor testing, some of the key focus areas are (1) understanding the complex thermal-hydraulic systems and materials that will facilitate heat removal from the reactor core, (2) mitigating the corrosion-associated issues that arise from using these materials at high temperatures, and (3) understanding how to measure and control salt composition/chemistry and properties during irradiation. This paper details the progress made toward surmounting these challenges to support future molten salt cartridge experiments in the VTR. Broadly, this work involves two major thrusts: design and analysis of an operating cartridge loop, and development of the instrumentation and control system needed to operate the loop successfully.

Preliminary MHD pressure drop analysis for the prototypical WCLL TBM with RELAP5/MOD3.3

In fusion reactor development, the design of the breeding blanket (BB) represents an important technical challenge due to the extreme operative conditions that the component must withstand. One of the critical issues of a liquid metal BB is the occurring of magnetohydrodynamic (MHD) effects within the piping network that transports the electro-conductive fluid. Predicting MHD phenomena is fundamental for a reliable and efficient blanket design and computational tools are indispensable to achieve a fusion-relevant physical analysis. For this reason, a dedicated MHD module for the System Thermal-Hydraulic (STH) code RELAP5/MOD3.3 (RELAP5) is under development at Sapienza University of Rome. In its current state, RELAP5 can evaluate MHD pressure losses in many common layouts for BB piping system. In a previous work of the Authors, the verification and validation (V&V) procedure of the MHD module has been discussed. In this paper, the code is used to assess the MHD pressure loss in the Water-Cooled Lithium Lead Test Blanket Module (TBM). Furthermore, the geometry of the TBM box is optimized to reduce the MHD pressure loss and to equalize the flow distribution among the Breeding Units.

A Hybrid Domain Overlapping Method for Coupling System Thermal Hydraulics and Cfd Codes

A Hybrid Domain Overlapping Method for Coupling System Thermal Hydraulics and Cfd Codes

Start-up, operation and thermal-hydraulic analysis of a self-propelling supercritical CO2 heat removal system coupled to a pressurized water reactor

The supercritical carbon dioxide (sCO2) heat removal system, which is based on a closed Brayton cycle with sCO2 as a working fluid, is an innovative, self-propelling and modular heat removal system for existing and future nuclear power plants. By changing the number of CO2 cycles, the heat removal capacity can be adapted. In this paper, up to four sCO2 cycles are analyzed in interaction with a pressurized water reactor, using the thermal-hydraulic system code ATHLET and considering a long-term station blackout and loss of ultimate heat sink scenario with conservatively high and low decay heat curves. The presented start-up procedure for the heat removal system might require further optimization due to the non-linear thermal gradients. Independent from the start-up, a heat removal system with three or four CO2 cycles keeps the primary loop temperatures sufficiently low. However, with only three cycles, the core is almost uncovered, and the danger of recriticality may occur due to cold leg deboration. Controlling the turbine inlet temperature via the turbomachinery speed and subsequent shutdown of single cycles successfully adapts the operation of the heat removal system to the declining decay heat. This enables reliable decay heat removal for more than 72 h.

Review and outlook of the integral test facility PKL III corresponding studies

Abstract This paper reviews an important integral test facility (ITF) named PKL (primary loop in German), which is designed based on a 4-loop pressurized water reactor (PWR) with the power 1 300 MWe, and especially concentrates on two aspects: (1) the tests at each developmental period of the facility until 2020, which is a typical microcosm of nuclear safety research; (2) the simulation of the PKL facility tests by using system thermal-hydraulic (STH) codes, especially RELAP5, TRACE and ATHLET. The results from the literature showed that all of these codes could reproduce the accident scenarios on the PKL facility to some extent, and simulate the complex phenomena both in the reactor pressurized vessel (RPV) and in the loops well, except some local phenomena (e. g., peak cladding temperature (PCT)). Furthermore, this paper presents some suggestions on PKL further tests. Especially, the sensitivity studies of initial conditions (ICs) and boundary conditions (BCs), test studies related to Extensive damage mitigation guidelines (EDMGs) and FLEX strategies, anticipated transients without scram (ATWS), detailed core section model, combination with other ITF or separate effects test (SET) facilities, and tests on advanced conception reactors are emphasized.

Simulation of VVER-1000 Guillotine Large Break Loss of Coolant Accident Using RELAP5/SCDAPSIM/MOD3.5

The safety performance of nuclear power plants (NPPs) is a very important factor in evaluating nuclear energy sustainability. Safety analysis of passive and active safety systems have a positive influence on reactor transient mitigation. One of the common transients is primary coolant leg rupture. This study focused on guillotine large break loss of coolant (LB-LOCA) in one of the reactor vessels, in which cold leg rupture occurred, after establishment of a steady-state condition for the VVER-1000. The reactor responses and performance of emergence core cooling systems (ECCSs) were investigated. The main safety margin considered during this simulation was to check the maximum value of the clad surface temperature, and it was then compared with the design licensing limit of 1474 K. The calculations of event progression used the engineering-level RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program, which also provide a more detailed treatment of coolant system thermal hydraulics and core behavior. The obtained results show that actuation of ECCSs at their actuation set points provided core cooling by injecting water into the reactor pressure vessel, as expected. The peak cladding temperature did not overpass the licensing limit during this LB-LOCA transient. The primary pressure above the core decreased rapidly from 15.7 MPa to 1 MPa in less than 10 s, then stabilizes up to the end of transient. The fuel temperature decreased from 847 K to 378 K during the first 30 s of the transient time. The coolant leakage reduced from 9945 kg/s to approximately 461 kg/s during the first 190 s in the transient. Overall, the study shows that, within the design of the VVER-1000, safety systems of the have inherent robustness of containing guillotine LB-LOCA.

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.

Partition method of wall friction and interfacial drag force model for horizontal two-phase flows

The improvement of thermal-hydraulic analysis techniques is essential to ensure the safety and reliability of nuclear power plants. The one-dimensional two-fluid model has been adopted in state-of-the-art thermal-hydraulic system codes. Current constitutive equations used in the system codes reach a mature level. Some exceptions are the partition method of wall friction in the momentum equation of the two-fluid model and the interfacial drag force model for a horizontal two-phase flow. This study is focused on deriving the partition method of wall friction in the momentum equation of the two-fluid model and modeling the interfacial drag force model for a horizontal bubbly flow. The one-dimensional momentum equation in the two-fluid model is derived from the local momentum equation. The derived one-dimensional momentum equation demonstrates that total wall friction should be apportioned to gas and liquid phases based on the phasic volume fraction, which is the same as that used in the SPACE code. The constitutive equations for the interfacial drag force are also identified. Based on the assessments, the Rassame-Hibiki correlation, Hibiki-Ishii correlation, Ishii-Zuber correlation, and Rassame-Hibiki correlation are recommended for computing the distribution parameter, interfacial area concentration, drag coefficient, and relative velocity covariance of a horizontal bubbly flow, respectively.

Study on LOFA and LOHS Accidents with Passive Safety System for Integrated Marine Reactor

To simulate the complex accident phenomena of a marine reactor, the thermal-hydraulic system code RELAP5 is modified to perform the analysis under ocean conditions. An integrated reactor with a passive residual heat removal system (PRHRS) is modeled by the improved code, and the effects of different ocean motions under a total loss-of-flow accident (LOFA) and a loss-of-heat-sink (LOHS) accident are analyzed with respect to safety characteristics. The results indicate that for LOFA, the primary loop can form an effective natural circulation to cool the core, and for LOHS, the PRHRS can effectively remove the residual heat from the core to ensure the core safety. The results also show that heaving motion accelerates the drop of the first-loop temperature and enhances the heat transfer capacity of the PRHRS. Inclining motion reduces the natural circulation flow in the core. A rolling condition causes fluctuations in the mass flow rate, the variations of which are not strictly sinusoidal, and increasing the rolling period also improves the heat exchange capacity of the PRHRS.

Superphénix Benchmark Part II: Transient Results

Abstract The paper presents a transient simulation phase of the new benchmark on a large sodium fast reactor (SFR). This phase of the benchmark is devoted to the modeling of selected operational transients performed during startup tests of the French SFR Superphénix. Six operational transients were selected for the analysis. The specifications of a simplified thermal hydraulic model equipped with point kinetics reactivity data and boundary conditions for the selected transients are given in the paper. The developed model contains a necessary thermal hydraulic description of the primary system components, assumptions to account for thermal expansion reactivity feedback from out-of-core structures, neutron kinetics parameters, power distribution, and reactivity coefficients. The neutronic input parameters were obtained with the help of the Monte Carlo code serpent during the first phase of the benchmark related to the static neutronic characterization of the core. In this study, the solution of the transient benchmark was obtained with three thermal hydraulic system codes, namely, TRACE, SIM-SFR, and ATHLET. The numerical results, compared to the available experimental data, exhibit a reasonable mutual agreement. Particular discrepancies between calculations and experiments could not be fully resolved. Therefore, a set of recommendations for achieving an improved agreement was proposed. In general, the proposed transient benchmark can be seen as an effective tool for validation and cross comparisons of system codes applied for safety analyses of SFRs, including approbation and comparison of different modeling features for thermal expansion of the out-of-core structures.

Using Machine Learning to Predict the Fuel Peak Cladding Temperature for a Large Break Loss of Coolant Accident

In this paper the use of machine learning (ML) is explored as an efficient tool for uncertainty quantification. A machine learning algorithm is developed to predict the peak cladding temperature (PCT) under the conditions of a large break loss of coolant accident given the various underlying uncertainties. The best estimate approach is used to simulate the thermal-hydraulic system of APR1400 large break loss of coolant accident (LBLOCA) scenario using the multidimensional reactor safety analysis code (MARS-KS) lumped parameter system code developed by Korea Atomic Energy Research Institute (KAERI). To generate the database necessary to train the ML model, a set of uncertainty parameters derived from the phenomena identification and ranking table (PIRT) is propagated through the thermal hydraulic model using the Dakota-MARS uncertainty quantification framework. The developed ML model uses the database created by the uncertainty quantification framework along with Keras library and Talos optimization to construct the artificial neural network (ANN). After learning and validation, the ML model can predict the peak cladding temperature (PCT) reasonably well with a mean squared error (MSE) of ∼0.002 and R2 of ∼0.9 with 9 to 11 key uncertain parameters. As a bounding accident scenario analysis of the LBLOCA case paves the way to using machine learning as a decision making tool for design extension conditions as well as severe accidents.

Application of artificial neural network for the critical flow prediction of discharge nozzle

System thermal-hydraulic (STH) code is adopted for nuclear safety analysis. The critical flow model (CFM) is significant for the accuracy of STH simulation. To overcome the defects of current CFMs (low precision or long calculation time), a CFM based on a genetic neural network (GNN) has been developed in this work. To build a powerful model, besides the critical mass flux, the critical pressure and critical quality were also considered in this model, which was seldom considered before. Comparing with the traditional homogeneous equilibrium model (HEM) and the Moody model, the GNN model can predict the critical mass flux with a higher accuracy (approximately 80% of results are within the ±20% error limit); comparing with the Leung model and the Shannak model for critical pressure prediction, the GNN model achieved the best results (more than 80% prediction results within the ±20% error limit). For the critical quality, similar precision is achieved. The GNN-based CFM in this work is meaningful for the STH code CFM development.

Assessing the effects of switching from multi-channel to three-dimensional nodalisations of the core in TRACE

Modern thermal-hydraulics (TH) systems analysis codes are increasingly being updated to include three-dimensional (3D) TH components towards capturing the complex flow phenomena in large components of light water reactors (LWRs). These are becoming increasingly popular for modelling reactor pressure vessels and cores since they potentially offer a more realistic representation of, for example, the effects of asymmetric loop flows and injection, which cannot be captured using traditional multi-channel models. Experience suggests, however, there may be unexpected side effects of upgrading from traditional multi-channel models. In this paper, we present comparative results demonstrating our experiences in switching to 3D components using the US-NRC system code TRACE. We present a series of studies in which 1D and 3D nodalisations are compared for representative PWR loss-of-coolant-accident (LOCA) scenarios. We see a clear trend for many of the LOCA scenarios considered. In particular, the peak cladding temperature (PCT) for 3D nodalisations is significantly lower than for traditional multi-channel models in many cases. Counter current flow limitation (CCFL) at the core upper tie plate appears to be one of the driving phenomena for many of the observed differences. This trend is, however, not universal; For a surge line break in the large scale test facility (LSTF), heterogeneities in the core upper tie plate lead to hot regions in the core, which are not captured in the multi-channel model. For large PWR cores, 3D nodalisations appear to represent the reality more closely. For smaller cross-sections, however, the validity of the TRACE’s CCFL prediction for 3D nodalisations is questionable. Specific guidelines are therefore needed to help TRACE users select the most appropriate 3D nodalisation for different geometries.

Development of a RELAP5/MOD3.3 Module for MHD Pressure Drop Analysis in Liquid Metals Loops: Verification and Validation

Magnetohydrodynamic (MHD) phenomena, due to the interaction between a magnetic field and a moving electro-conductive fluid, are crucial for the design of magnetic-confinement fusion reactors and, specifically, for the design of the breeding blanket concepts that adopt liquid metals (LMs) as working fluids. Computational tools are employed to lead fusion-relevant physical analysis, but a dedicated MHD code able to simulate all the phenomena involved in a blanket is still not available and there is a dearth of systems code featuring MHD modelling capabilities. In this paper, models to predict both 2D and 3D MHD pressure drop, derived by experimental and numerical works, have been implemented in the thermal-hydraulic system code RELAP5/MOD3.3 (RELAP5). The verification and validation procedure of the MHD module involves the comparison of the results obtained by the code with those of direct numerical simulation tools and data obtained by experimental works. As relevant examples, RELAP5 is used to recreate the results obtained by the analysis of two test blanket modules: Lithium Lead Ceramic Breeder and Helium-Cooled Lithium Lead. The novel MHD subroutines are proven reliable in the prediction of the pressure drop for both simple and complex geometries related to LM circuits at high magnetic field intensity (error range ±10%).

Global Sensitivity Analysis and Bayesian Calibration on a Series of Reflood Experiments with Varying Boundary Conditions

Best estimate plus uncertainty for the safety assessment of nuclear power plant transient requires, among others, estimating the probability density function (PDF) of physical model parameters in thermal-hydraulic system codes. In that context, Bayesian calibration based on experimental data from separate-effect test facilities are increasingly popular to inform the PDF of a single thermal-hydraulic phenomenon. These calibrations are, however, time intensive, especially when considering multiple time-dependent outputs. Calibrating on many tests with different boundary conditions and potentially different phenomena to derive PDFs applicable to complex transients appears intractable, even using hierarchical modeling. In this paper, we start investigating this problem by considering a set of Flooding Experiments with Blocked Arrays reflood tests with different boundary conditions. We use TRACE v5.0p3 to model time- and space-dependent temperature profiles, pressure drops, and liquid carry-over. Global sensitivity analysis helps screen out noninfluential parameters and gain a detailed understanding of the modeled physics of reflood. The analysis shows that, for all tests, the outputs were sensitive to a similar set of influential model parameters. In turn, Bayesian calibration yields similar posterior PDFs for the influential parameters, and forward propagation of these posterior PDFs yields similar confidence intervals. As such, the information of the investigated tests can well be represented by a unique posterior PDF. Such simplifications, although not general, are welcome to help manage the intensive calibration effort necessary for dealing with complex thermal-hydraulic transients encountered in nuclear power plants.

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.

Simulation of Partial and Complete Loss of Flow Accidents for GPWR using ATHLET Code

In Nuclear power plants, simulation of the thermal hydraulic systems at loss of flow, accidents is commonly used to perform safety and licensing analyses through regulatory and safety authorities since they are considered as design base accidents. In this paper, a simulation of a thermal-hydraulic model of Germany Pressurized Water Reactor (GPWR) is developed, during loss of flow accident (LOFA) scenarios, such as: trip of one, two, three and four of the Main Coolant Pumps, at operation of 100% reactor power. The developed thermal hydraulic model is formed using ATHLET3.1A code with a point reactor kinetic to predict the behavior of the reactor through the thermal hydraulic major parameters such as power, mass flow rate, primary and secondary pressures, max clad surface and fuel temperatures. All reactivity feedbacks are taken into consideration in the developed simulation and the scram signal of the protection system depends on the mass flow rate. The results show that the investigated ATHLET model can precisely simulate, the steady state and the transient behavior of GPWR during many LOFA accidents. Moreover, the results show that there is good agreement between ATHLET model and RELAP5 code results are nearly in agreement with in previous literatures

Numerical study on thermal-hydraulics of external reactor vessel cooling in high-power reactor using MARS-KS1.5 code: CFD-aided estimation of natural circulation flow rate

This paper presents a numerical investigation of two-phase natural circulation flows established when external reactor vessel cooling is applied to a severe accident of the APR1400 reactor for the in-vessel retention of the core melt. The coolability limit due to external reactor vessel cooling is associated with the natural circulation flow rate around the lower head of the reactor vessel. For an elaborate prediction of the natural circulation flow rate using a thermal-hydraulic system code, MARS-KS1.5, a three-dimensional computational fluid dynamics (CFD) simulation is conducted to estimate the flow rate and pressure distribution of a liquid-state coolant at the brink of significant void generation. The CFD calculation results are used to determine the loss coefficient at major flow junctions, where substantial pressure losses are expected, in the nodalization scheme of the MARS-KS code such that the single-phase flow rate is the same as that predicted via CFD simulations. Subsequently, the MARS-KS analysis is performed for the two-phase natural circulation regime, and the transient behavior of the main thermal-hydraulic variables is investigated.

Qualification of RELAP5-3D code condensation model against full-scale PERSEO Test 9

Passive systems are being implemented in several nuclear plants already constructed, under construction or under design. The purpose of these systems is to increase the inherent safety of the plant in any transient condition, so they can provide a higher safety level in case of an accident. Best estimate thermal-hydraulic system codes, widely adopted in the nuclear field to perform deterministic safety analyses, need to be validated against the phenomena typical of passive systems (e.g. natural circulation, thermal-hydraulic behavior of large pools, low pressure phenomena, etc.). An experimental test facility called PERSEO (in-Pool Energy Removal System for Emergency Operation) was built at SIET laboratory in Piacenza (Italy) modifying the existing PANTHERS IC-PCC facility. It is a full-scale test facility aimed at studying a new passive heat removal system and it can be used to analyze the capability of system codes to simulate the related phenomena. In the present analysis, Test 9, which was performed in the PERSEO facility at around 4.1 MPa to study the activation and the heat removal capability of the proposed system, has been simulated with the RELAP5-3D code to assess its capability to predict the dominant phenomena of the transient. Particular attention has been dedicated to the analysis of condensation heat transfer coefficient computed by the code, since in previous analyses it has been found an evident underestimation of the power exchanged by the passive system. Alternative correlations for the condensation heat transfer coefficient have been identified and a correction factor for the heat transfer coefficient calculated by RELAP5-3D has been computed. After the application of the correction factor, the set composed by Nusselt-Kutateladze-Chen correlations provided the better agreement with the experimental data. Finally, the accuracy of the code calculation has been evaluated qualitatively and quantitatively by adopting the Fast Fourier Transform Based Method.