TRAC/RELAP Advanced Computational Engine Research Articles

Validation of TRACE 5.0 for steam line break with passive auxiliary feedwater system via the ATLAS facility

The TRAC/RELAP Advanced Computational Engine (TRACE) reactor systems code was validated for the Passive Auxiliary Feedwater System (PAFS) using the experimental data from the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) facility. The accident scenario which emulates a Double-Ended Guillotine Break (DEGB) in the main steam line of Steam Generator 1, was meticulously analyzed, with a focus on transient events and steady-state conditions with an error margin of less than 2%. Comprehensive comparative analyses of key thermal hydraulic parameters, including system pressures, temperatures, flow rates, and water levels, were conducted. Discrepancies were systematically identified, investigated, and effectively resolved. The experimental results underscore the successful role of passive safety systems in mitigating potential accident impacts, emphasizing their pivotal contribution to the overall safety and reliability of nuclear reactor systems. Furthermore, the simulation results effectively mirror the experimental data, demonstrating TRACE's capability to adequately simulate the behavior of a passive feedwater system during a main steam line break.

BEPU analysis of an Upgraded ICE facility test by TRACE/DAKOTA coupling

The Ingress of Coolant Event (ICE) in the plasma chamber is one of the safety issues in fusion nuclear plants. The best estimate thermal-hydraulic system codes adopted to perform deterministic safety analysis should be validated against the phenomena typical of accidental transients in fusion installations. TRACE (TRAC/RELAP Advanced Computational Engine), best estimate thermal-hydraulic system code developed by USNRC, has been adopted to simulate an ICE. The calculated results have been compared to the experimental data obtained in one test performed in the upgraded Integrated ICE facility at JAERI. In this updated configuration the pressure suppression system is connected to the top of the plasma chamber instead of the bottom of the vacuum vessel. The facility nodalization has been developed in the SNAP environment/architecture. To qualify the code and the nodalization, an accuracy evaluation has been performed both from a qualitative and quantitative point of view. Then, considering the presence of some uncertainties in the input-deck development, an uncertainty analysis has been carried out. The probabilistic method to propagate the input uncertainties has been selected and the analysis has been carried out with the DAKOTA toolkit coupled with TRACE code in SNAP. In the uncertainty analysis, some relevant statistical parameters have been considered to characterize the dispersion of the results and the correlation between the uncertain input parameters selected and the PC pressure chosen as figure of merit.

Application of REPAS to analyze the sump clogging issue following a LOCA and its impact on the reliability of the ECCS long-term core cooling function

Sump clogging has been identified as a relevant issue after an accident occurred in the Barsebäck-2 nuclear power plant in Sweden (1992). Following a steam line Loss of Coolant Accident (LOCA) due to the inadvertent opening of a safety relief valve, the jet stripped some insulation material from nearby pipes. The insulation debris were transported to the inlet of the strainers for the drywell spray system, thus clogging the intake. The accident was not serious but showed that the Emergency Core Cooling System (ECCS) could have failed. As a consequence, several actions have been undertaken by international organizations, regulatory bodies and nuclear power plant owners to characterize this issue, and propose solutions and improvements. In the present paper, the Reliability Evaluation of Passive Safety Systems (REPAS) methodology is applied to analyze the sump clogging issue and its effect on the long-term core cooling function. Originally developed to evaluate the reliability of passive systems, REPAS is here applied for the first time to an active system. The application is performed using the TRAC/RELAP Advanced Computational Engine (TRACE) code, developed by the USNRC, to simulate a generic three-loop PWR, with an active decay heat removal system.

Analysis of LBLOCA of APR1400 with 3D RPV model using TRACE

It is very difficult to capture the multi-dimensional phenomena such as asymmetric flow and temperature distributions with the one-dimensional (1D) model, obviously, due to its inherent limitation. In order to overcome such a limitation of the 1D representation, many state-of-the-art system codes have equipped a three-dimensional (3D) component for multi-dimensional analysis capability. In this study, a standard multi-dimensional analysis model of APR1400 (Advanced Power Reactor 1400) has been developed using TRACE (TRAC/RELAP Advanced Computational Engine). The entire reactor pressure vessel (RPV) of APR1400 has been modeled using a single 3D component. The fuels in the reactor core have been described with detailed and coarse representations, respectively, to figure out the impact of the fuel description. Using both 3D RPV models, a comparative analysis has been performed postulating a double-ended guillotine break at a cold leg. Based on the results of comparative analysis, it is revealed that both models show no significant difference in general plant behavior and the model with coarse fuel model could be used for faster transient analysis without reactor kinetics coupling. The analysis indicates that the asymmetric temperature and flow distributions are captured during the transient, and such nonuniform distributions contribute to asymmetric quenching behaviors during blowdown and reflood phases. Such asymmetries are directly connected to the figure of merits in the LBLOCA analysis. Therefore, it is recommended to employ a multi-dimensional RPV model with a detailed fuel description for a realistic safety analysis with the consideration of the spatial configuration of the reactor core.

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

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

Simulation of natural circulation cartridge loop experiments and application to molten salt reactors

This work uses the TRAC/RELAP Advanced Computational Engine (TRACE) thermal hydraulics (TH) code to model natural circulation cartridge loop experiments previously conducted at Oak Ridge National Laboratory (ORNL) using water and compares the simulated and experimental results. TRACE is also used to characterize natural circulation in the cartridge loop vehicle using FLiNaK as the working fluid. The experimental vehicle is a buoyancy-aided, annular cartridge loop, referred to as a thermosyphon, and is designed to aid in qualifying liquid–fueled and/or liquid–cooled irradiation experiments for the Versatile Test Reactor (VTR), which is currently being designed in the United States. Out-of-pile water experiments have been conducted using the cartridge and the Thermosyphon Test Loop facility at ORNL, and future experiments are anticipated that would use other molten salt surrogates as the working fluid, followed by eventual insertion of a similar cartridge into VTR. This work aims to determine how well TRACE can replicate the natural convection conditions that were observed experimentally; this serves as an initial step for validating the modeling tool for design and safety calculations to support future irradiation experiments in VTR. Initial predictions of potential experiments were made using FLiNaK as the natural circulation fluid to demonstrate the relevance of the cartridge design to molten salt reactors (MSRs). Results from this study indicate that TRACE can accurately capture natural convection phenomena in the thermosyphon and that several design changes to the current cartridge vehicle are necessary to achieve hydraulic conditions similar to those expected in MSRs.

Experimental and Numerical Studies of Pressure and Oscillatory Flow Effects on the Two-Phase Flow and Heat Transfer in a Rod Bundle

An experimental and numerical study was performed to investigate the effects of pressure and constant vs oscillatory flooding rates on the two-phase flow and heat transfer behavior of a rod bundle under reflood transient conditions. Experimental results were obtained from the NRC/PSU Rod Bundle Heat Transfer (RBHT) test facility from various test cases covering a range of system pressures with light water as the working coolant. For each pressure case, two experiments were performed, one for a constant flow rate, and one for an oscillating flow rate about the constant flow rate. The RBHT test facility, which contains 49 vertical, 3.66 m (12 ft) long test rods (four unheated corner rods and 45 heated rods) with Inconel 600 cladding in a 7 × 7 geometry, having the rod diameters, rod pitches and spacer grids comparable to those in commercial PWRs, was specifically designed to obtain fundamental flow and heat transfer data during reflood transients. The thermal–hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) was used in this study by performing simulations with the same geometry and operating conditions as the RBHT facility for each experiment. Results of the TRACE simulations were compared to the experimental data obtained in the RBHT tests. It was found that the trends on the pressure effects for constant and oscilatory flows on the thermal–hydraulic behavior of the rod bundle (i.e., the two-phase flow and heat transfer behavior of the rod bundle during reflood transients) predicted by the TRACE model agree well with the RBHT data. This comparison of results has also assisted in other studies to investigate numerical discrepancies currently underway.

The evaluation of TRACE/PARCS model for BWR-4 nuclear power plant by startup test transient analyses

Abstract Generally, the thermal hydraulic (TH) codes need the results of Neutron Kinetics (NK) codes providing the reactivity properties to calculate neutron flux. Then the TH codes perform the safety analyses obtaining the responses of pressure, temperature, or water level. Two kinds of different codes calculate different physical behaviors sequentially and separately. Simultaneously computing thermal hydraulic and neutron kinetics behaviors can enhance the accuracy of the analysis. Hence, it is crucial to develop the TH-NK coupled model. This study presents the capability of the TH-NK coupled model, developed by TRACE (TRAC/RELAP Advanced Computational Engine) and PARCS (Purdue Advanced Reactor Core Simulator), for the BWR-4 nuclear power plant. The establishment of the TRACE/PARCS model presented the nodal and component modeling methodologies. This model was used to simulate two startup tests of high power level system transients. Principal system responses, calculated by the TRACE/PARCS model, were compared with the measured data in startup tests and the results of the point kinetic calculation of the TRACE (TRACE/PK) to evaluate the model. The evaluation shows that the TRACE/PARCS model can simulate the interaction between thermal hydraulic and neutron kinetics phenomena and predict the transients suitably. Through the comparison, the TRACE/PARCS model can be confident doing the analyses of normal and abnormal operational transients to predict the transient responses.

Analysis of nodalization strategies to model a suppression tank for fusion plants with TRACE code

The Ingress of Coolant Event (ICE) in the plasma chamber is one of the main safety issues in nuclear fusion plants. The ICE is caused by the rupture of the coolant tubes installed in the plasma facing components; the coolant ingress causes a pressure rise in the plasma chamber and vacuum vessel, normally under high-vacuum conditions. To mitigate the system pressurization leading to mechanical structure failure, a pressure suppression system is installed. Safety analyses of the hypothetical challenging accidental scenarios can be conducted by deterministic models that need to be validated against experimental data characterizing the target phenomena. The paper presents a study of different nodalization strategies for modeling a suppression tank for fusion plants, using the best estimate thermal-hydraulic system code TRACE (TRAC/RELAP Advanced Computational Engine). The TRACE code is developed by USNRC to perform safety analyses for light water fission reactors. Both mono-dimensional and three-dimensional approaches were adopted to model the suppression tank. The experimental data from the JAERI Integrated ICE facility (scaling factor of 1/1600 with respect to the ITER-FEAT design) was used to benchmark different nodalization options. In addition to the qualitative accuracy evaluation, the Fast Fourier Transform Based Method (FFTBM) was applied to quantify the accuracy of the code results for different nodalizations.

Development and verification of the coupled thermal–hydraulic code – TRACE/SCF based on the ICoCo interface and the SALOME platform

The system thermal–hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) and the sub-channel code SubChanFlow (SCF) have been coupled together based on the Interface for Code Coupling (ICoCo) and the SALOME platform. It is a server-client system where TRACE and SCF are computational engines, ICoCo transfers the coupling data, and SALOME supervises the two clients’ synchronization. TRACE and SCF run serially. The coupled code adopts the domain-overlapping (assisted by a Dynamic Implicit Additional Source (DIAS) method) and the Operator Split (OS) explicit coupling approaches. The MEDCoupling library manages the field mapping between different meshes of TRACE and SCF. The YACS module in SALOME provides developers with useful tools that can implement numerical codes in SALOME. It provides users with Graphic User Interfaces (GUI) that can construct and execute the coupling simulations as well. Users can conveniently build/optimize/execute the coupling schemes in SALOME. The coupled code has been verified with an AP1000 feedwater-temperature-drop transient.

Full-scale dryout/rewet instability tests in a BWR rod bundle and TRACE assessment

The regulatory purpose of the current work is to study fuel damage mechanisms and processes that were predicted to occur under certain unstable conditions that could arise during postulated anticipated transients without SCRAM (ATWS) for boiling water reactors (BWRs) operating in the maximum extended load line limit analysis plus (MELLLA+) expanded operating domain. The Nuclear Regulatory Commission (NRC) conducted experiments at the high pressure Karlstein Thermal-Hydraulic Experimental Facility (KATHY) using a full-scale simulated BWR fuel assembly. The simulated assembly includes representative features of modern BWR fuel bundles, including commercial spacers, part length fuel rods and water rods. KATHY includes two features that allowed the NRC to conduct experiments that subject the simulated fuel assembly to unstable conditions similar to those expected during postulated ATWS events. These features are the downcomer, which allows the loop to be operated under natural circulation flow, and the proprietary SINAN control program, which allows the heater rod voltage to vary in response to changing thermal-hydraulic conditions, which in turn, simulates reactivity feedback. The NRC conducted several tests which produced growing flow and power oscillations and these oscillations were allowed to proceed up to the point where the heater rod surface failed to rewet and went into the early stages of heat-up. The experimental data were analyzed to determine the temperature threshold for failure to rewet and the experimental results have been compared to various models of the minimum stable film boiling temperature (Tmin). The results show that the homogeneous nucleation temperature provides a reasonable estimate of this temperature threshold. Further, the NRC's TRAC-RELAP Advanced Computational Engine (TRACE) code was used to simulate the experiments to assess the TRACE performance for simulating failure to rewet during postulated ATWS-I events. When compared to the experimental data, TRACE produces slightly conservative estimates of the cladding temperature when an appropriately low Tmin value is used (i.e., one consistent with homogeneous nucleation). Therefore, TRACE is appropriate for predicting cladding heat-up during postulated ATWS-I events.

TRACE Analysis of a Loss of Alternating-Current Power Without Rod Insertion for the NuScale Power Module—I: Basic Event Progression and Long-Term Behavior

The U.S. Nuclear Regulatory Commission (NRC) staff often performs confirmatory analysis to support regulatory decision making. In the current work the TRAC/RELAP Advanced Computational Engine (TRACE) code was used to study the transient system response for the NuScale power module to a beyond-design-basis event where the control rods fail to insert. The regulatory purpose of the current work was to confirm the results of analyses provided by the applicant as part of their probabilistic risk assessment analysis that demonstrates that the core is not damaged under certain conditions when the control rods fail to insert. The NRC staff performed calculations using a TRACE model of the NuScale power module that includes both the primary and secondary systems that simulates a loss of alternating-current power and complete failure of the module protection system to insert control rods. The NRC staff analyses demonstrate that under these conditions the reactor stabilizes at a new stable condition with minor power and pressure oscillations where core power is balanced by passive heat removal.

TRACE Analysis of a Loss of Alternating-Current Power Without Rod Insertion for the NuScale Power Module—II: Sensitivity to Varying Initial Temperature

In a companion paper, the U.S. Nuclear Regulatory Commission (NRC) staff has described analyses performed using the TRAC/RELAP Advanced Computational Engine (TRACE) code to study the transient system response of the NuScale power module to a postulated beyond-design-basis loss of alternating-current (LOAC) power transient where the module protection system completely fails to insert the control rods. The subject paper studies the sensitivity of the event progression and consequences to variation in the initial reactor coolant system (RCS) temperature. These studies were performed by varying the effective steam generator heat transfer surface area between 100% and 50% of the nominal area. The results of the NRC staff analyses show that at increased initial temperatures, it is possible for the NuScale primary side to remain critical for an extended period of time, leading to a sustained loss of primary-side inventory through pressure relief until the natural circulation flow pattern in the RCS becomes broken. After the flow loop is broken, reactor power decreases significantly, and the primary figures of merit important to safety are met with substantial margin.

A Rigorous Method for Selecting the Most Limiting Exposure During Cycle for Anticipated Transient Without SCRAM with Instability Analysis

The U.S. Nuclear Regulatory Commission (NRC) staff often perform confirmatory analyses using the TRAC/RELAP Advanced Computational Engine (TRACE) and Purdue Advanced Reactor Core Simulator (PARCS) codes to assist in regulatory decision making. Recently, the NRC staff have performed numerous such analyses of anticipated transient without SCRAM (ATWS) with core instability (ATWS-I) scenarios for boiling water reactor license amendment requests to expand the power/flow operating domain. In the conduct of these confirmatory analyses, the staff have simulated oscillatory conditions in the reactor core under certain ATWS conditions that result in regional mode (or out-of-phase mode) power oscillations. The nature of these regional oscillations may present a challenge to fuel damage limits. Therefore, there has been interest in methods to identify the most limiting point in cycle exposure. It has been conventional wisdom that the core is most susceptible to regional mode oscillations when the fission cross section is greatest, leading to the common practice of analyzing these events at the peak hot excess (PHE) exposure point in the cycle. The staff have found some limitations in applying the PHE concept in a consistent manner. In the current work, the NRC staff have developed a more rigorous method for identifying the most limiting cycle exposure by directly considering the core flow rate, the axial power distribution, the first harmonic mode shape, and the eigenvalue separation between the fundamental and first harmonic modes. This method is a more rigorous method to screen the various exposures between beginning and end of cycle. An example case is shown to demonstrate the application of this methodology.

Validation and Uncertainty Quantification for Two-Phase Natural Circulation Flows Using TRACE Code

The work presents validation of the TRAC/RELAP Advanced Computational Engine (TRACE) code for natural circulation two-phase flow in a vertical annulus. Natural circulation experiments were recently conducted for a vertical internally heated annulus at the Multiphase Thermo-Fluid Dynamics Laboratory at the University of Illinois. The experimental matrix consists of 107 experiments with system pressure in the range of 145 to 950 kPa and heat flux up to 275 kW/m2. Void fraction, gas velocity, and interfacial area concentration were measured in five axial locations along the test section with six measurements of bulk liquid temperature and pressure. To validate the capability of the TRACE code under natural circulation flow conditions, a complete model of the experimental facility was created and validated using forced convection and single-phase natural circulation data. Sensitivity and uncertainty quantification were performed. The sensitivity to important simulation parameters was studied using Sobol’s variance decomposition and the Morris screening method. The sensitivity of boundary conditions on void fraction measurement was investigated. The sensitivity study has shown significant differences in model sensitivity between different experimental conditions. With heat flux being the most influential parameter for high-pressure cases without flashing and pressure, temperature and heat flux have a combined strong effect in the case of low-pressure experiments when flashing occurs. Additionally, higher uncertainty in void fraction prediction was observed for experimental conditions at low pressure with flashing.

Comparison of water, helium, and carbon dioxide as coolants for next generation power plants using TRACE

The purpose of this study is to compare the use of water, helium, and carbon dioxide as coolants for Generation IV and fusion power plants. In this study, the United States Nuclear Regulatory Commission’s (NRC’s) thermal hydraulics and neutronics coupled code TRAC/Relap Advanced Computational Engine (TRACE) was used to build and simulate a 600 MWth fusion Field Reversed Configuration (FRC) system that was cooled by either water, helium, or carbon dioxide. The results from the simulations, along with certain design criteria, were used to determine that water was the best coolant for the system of this study. For the operating conditions used in this research, water was able to keep the temperature of certain materials below their maximum temperatures much easier than the helium and carbon dioxide cooled systems. Specifically, beryllium was used as a material in the system and was determined to have a maximum temperature of 800 °C for its applications in this study (Mitteau et al., 2017). Additionally, this temperature limit restricts the efficiency and capabilities of the helium and carbon dioxide cooled systems. There is also uncertainty associated with the turbine efficiencies used in this study. This introduces uncertainty into the overall efficiency of the system for each coolant. So, the overall efficiency is not as important of a parameter in determining the best coolant for the system in this study as other parameters.

Analyses of Feedwater Trip With SBO Sequence of VVER1000 Reactor

In the Czech Republic, as a follow-up, a consortium of research organizations and universities has decided to simulate selected stress tests’ scenarios, in station blackout (SBO) and the loss of ultimate heat sink (LoUHS), with the aim to verify the national stress report and to analyze time response of respective source term releases. These activities are carried out in the frame of the project prevention, preparedness, and mitigation of consequences of severe accident (SA) at Czech NPPs in relation to lessons learned from stress tests after Fukushima, financed by the Ministry of Interior. The Research Centre Rez has been working on a methods for estimation of leakages and consequences of releases (MELCOR) model for VVER1000 nuclear power plant (NPP) starting with a plant systems nodalization. The aim was to benchmark the MELCOR model with the validated TRAC/RELAP advanced computational engine (TRACE) model, first comparing the steady state and continuing in a long-term SBO plus another event until the beginning of the SA. The presented work is based on the previous paper from the ICONE 23rd Conference hosted in Japan. It focuses mainly on the comparison of the thermohydraulics of the two models created in MELCOR and TRACE codes as outcome of the “Fukushima project.” After that, preliminary general results of the SA progression showing the hydrogen production and the relocation phenomena will be shortly discussed. This scenario is considered closed after some seconds to the break of the lower head. It is important to note that this paper is a substantial update of a previous one (Mazzini et al., 2015, “Analyses of SBO Sequence of VVER1000 Reactor Using TRACE and MELCOR Codes,” 23rd International Conference on Nuclear Engineering (ICONE23)) and although it contains the same descriptive sections, all the results are new.

Emergency operating procedures improvement based on the lesson learned from the Fukushima Daiichi accident

One of the lessons learned from the Fukushima Daiichi accident is the emergency operating procedures (EOPs) have to be improved. The BWR Owners’ Group revised the emergency procedure guidelines and addressed the lesson learned from the Fukushima Daiichi accident in revision 3 in order to avoid loss of turbine-driven makeup water systems during reactor depressurization. However, the improvement deserves much more attention. The existing EOPs at the time of the accident may not be adequate enough for the prolonged station blackout condition, because resources required for performing the EOPs are vastly unavailable or gradually exhausted. The improved EOPs must not only permit early reactor pressure vessel depressurization, but also address the risk accompanied with the emergency depressurization. For this reason, Taiwan Power Company proposed the Ultimate Response Guideline (URG) to cope with Fukushima-like accidents. The main content of the URG is a two-stage depressurization strategy, namely the controlled depressurization and the emergency depressurization. The technical basis of the two-stage depressurization strategy was discussed in this paper. The effectiveness of the URG was verified by using TRAC/RELAP Advanced Computational Engine (TRACE). Besides, the emergency responses performed by Fukushima Daini nuclear power plant (Fukushima Daini NPP) were found to be very similar to the URG. The consequences of Fukushima Daini NPP somehow demonstrate that the URG is effective for Fukushima-like situation. Effective strategies for coping with Fukushima-like accidents are worthwhile to be integrated into the improved EOPs. The Fukushima Daiichi accident might have been prevented if the improved EOPs had been used, even with the existing design of Fukushima Daiichi NPP at the time of the accident.

Thermal hydraulics analysis of the Advanced High Temperature Reactor

The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

The development and assessment of TRACE/PARCS model for Lungmen ABWR

This study consists of two steps. The first step is the development of a TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model of Lungmen nuclear power plant (NPP) which includes the vessel, reactor internal pumps (RIPs), main steam lines, and important control systems (such as the feedwater control system, steam bypass & pressure control system, and recirculation flow control system), etc. Key parameters were identified to refine the TRACE/PARCS model further in the frame of a steady state analysis. The second step is the performance of Lungmen NPP TRACE/PARCS model transient analyses. The transient data of Final Safety Analysis Report (FSAR) chapter 15th were used to compare with the results of Lungmen NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the FSAR data. It indicated that there was a respectable accuracy in the Lungmen NPP TRACE/PARCS model and it also depicted that the Lungmen NPP TRACE/PARCS model was satisfying for the purpose of Lungmen NPP safety analyses.