System Code ATHLET Research Articles

Design and analysis on the HP-PHRS for small modular lead-bismuth fast reactor

The heat pipe is an efficient heat-conducting element that can efficiently and passively transfer heat from a high-temperature heat source to a low-temperature heat source. It is widely used in the various industries. Based on high-temperature heat pipe technology, this paper designs a passive decay heat removal system for a small modular lead–bismuth fast reactor, and uses the system code ATHLET coupled with the heat pipe calculation module to calculate and analyze the thermal–hydraulic characteristics of the reactor under station black out accident.

Validation of coupled ATHLET-OpenFOAM simulation on a large-scale single- and two-phase flow experiment

Abstract System thermal-hydraulics (STH) codes are successfully used in the last decades for design and analysis of the physical behavior of nuclear power plants (NPPs). These programs provide reliable results at comparatively low computational cost, but have limited capabilities to predict complex 3D flow phenomena, which is possible with Computational Fluid Dynamics (CFD) tools. In order to combine the advantages of STH and CFD codes, algorithms to couple these are developed. At GRS, a new OpenFOAM solver, based on the volume of fluid (VOF) method, was developed together with new coupling boundary conditions to couple the CFD program OpenFOAM with the system code ATHLET. In a next step, verification and validation activities were carried out. These included the analysis of a Pressurized Thermal Shock (PTS) scenario. For this purpose, data from the OECD/NEA ROSA V Test 1.1, carried out in the Japanese Large-Scale Test Facility (LSTF) was used. The experiment investigated the temperature stratification after emergency core cooling (ECC) injection under single- and two-phase natural circulation conditions in the primary cooling circuit of a pressurized water reactor (PWR). This paper discusses the results of the coupled simulations, compares the numerical results with experimental data, and identifies areas for future improvements.

Implementation of the preCICE coupling interface for AC2/ATHLET

Abstract The design of new nuclear reactor types as well as the analysis of certain phenomena in existing reactors require to consider different physical models and three-dimensional phenomena like the effect of turbulence in fluids and three-dimensional (3D) heat conduction in complex structures. One-dimensional (1D) lumped parameter system codes used to simulate transients in nuclear reactors lack high-resolution models. Special codes like computational fluid dynamics (CFD) or computational structural mechanics codes (CSM) can simulate those, but they require significantly more computational resources. Therefore, different coupling methods have been developed to limit the use of computationally expensive codes to those parts of a plant where they are really needed and couple them with systems codes, which simulate the rest. The library preCICE enables simultaneous coupling of multiple simulation programs, e.g., the 3D CFD code OpenFOAM and the 3D CSM code CalculiX. preCICE coupling interfaces were developed for the system code ATHLET. We focused on fluid-fluid couplings and conjugate heat-transfer couplings. Coupled simulations were performed for an experiment at the test facility TALL for the transition from forced to natural circulation coupling ATHLET with OpenFOAM, and a generic building condenser geometry coupling the three codes ATHLET, CalculiX, and OpenFOAM. The correct integral transfer of the quantities at the coupling interfaces was verified. However, it was found that coupled quantities at the interfaces need to be converted from 0D to 2D and backwards considering the underlying physics, which goes beyond the pure numerical considerations of the preCICE library.

Transient simulation and analysis of a supercritical CO2 heat removal system under different abnormal operation conditions

The supercritical carbon dioxide (sCO2) heat removal system, which is based on multiple closed Brayton cycles with sCO2 as the working fluid, is an innovative, self-propelling and modular heat removal system for existing and future nuclear power plants. Previous studies analysed its design, layout, control and operation. In addition, this study considers different sudden failures during the accident progress, e.g. failure of single sCO2 cycles, control systems and valves. These abnormal conditions were investigated with the thermal-hydraulic system code ATHLET for a generic Konvoi pressurized water reactor. In most cases, the failure of a single sCO2 cycle can be compensated. On the one hand, failure of the fans of the gas cooler leads to a pressure increase which may be mitigated by an inventory control system or cycle shutdown. On the other hand, unintended fan speed-up can cause compressor surge without adequate countermeasures. Furthermore, the system can operate under the cyclic blow-off from the steam generator safety valves when the relief valves are not available. Finally, the unintended closure of the valve which controls the steam flow through the compact heat exchanger triggers a fast cycle shutdown but a subsequent restart might be possible.

Extension of the DYN3D/ATHLET code system to SFR applications: models description and initial validation

DYN3D is a 3D reactor dynamics code originally developed for Light Water Reactors (LWRs) analyses. In recent years, the applicability range of DYN3D was extended to coupled 3D neutronics/thermal-hydraulics (TH) simulations of steady-states and transients in Sodium cooled Fast Reactors (SFRs) at reactor core level.The main objective of this study is to scale up the simulation capabilities of DYN3D from SFR core to system level by coupling DYN3D with a TH system code ATHLET capable of sodium flow modeling. The paper describes the adaption and extension of the existing LWR-oriented DYN3D/ATHLET coupling to SFR analysis. This includes a description of the coupling techniques used to integrate DYN3D with ATHLET and a summary of modifications required to enable the modeling of SFR specific phenomena. The paper also presents the approach to the modeling of reactivity effects caused by the thermal expansions of the reactor components located outside the reactor core, such as reactor vessel, core support structure (strongback), control rod drive lines (CRDLs), etc. The paper also includes the description of the initial verification and validation activities of the extended DYN3D/ATHLET code system. The unprotected stage of the Phenix End-Of-Life natural convection test is used as a test case for this purpose.

On the validation of ATHLET 3-D features for the simulation of multidimensional flows in horizontal geometries under single-phase subcooled conditions

This paper provides an assessment of fluid transport and mixing processes inside the primary circuit of the test facility ROCOM through the numerical simulation of Test 2.1 with the system code ATHLET. The experiment represents an asymmetric injection of cold and non-borated water into the reactor coolant system (RCS) of a pressurized water reactor (PWR) to restore core cooling, an emergency procedure which may subsequently trigger a core re-criticality. The injection takes place at low velocity under single-phase subcooled conditions and presents a major challenge for the simulation in lumped parameter codes, due to multidimensional effects in horizontal piping and vessel arising from density gradients and gravity forces. Aiming at further validating ATHLET 3-D capabilities against horizontal geometries, the experiment conditions are applied to a ROCOM model, which includes a newly developed horizontal pipe object to enhance code prediction inside coolant loops. The obtained results show code strong simulation capabilities to represent multidimensional flows. Enhanced prediction is observed at the vessel inlet compared to traditional 1-D approach, whereas mixing overprediction from the descending denser plume is observed at the upper-half downcomer region, which leads to eventual deviations at the core inlet.

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.

Simulation of steam/air leakage through small-scale cracked concrete specimen

At the Material Testing and Research Institute (MPA Karlsruhe) leakage experiments with concrete specimens containing a single penetrating crack were performed on behalf of GRS. The specimens had a prismatic shape and the dimensions 4.0 cm × 4.0 cm × 11.0 cm. The crack width was adjustable and ranged from 0.05 mm to 0.25 mm in the experiments. On top of the leakage measurement two sets of additional measurements were performed. In leakage experiments with air at room temperature the exit velocity of the leakage flow was measured with a hot-wire anemometer. The measured velocities were compared to the velocities calculated by leakage rate formulas. The formulas compare well to measurement results. Differences depend greatly on the crack width and the respective assumptions on the friction coefficient proposed by the authors. In a second set of experiments, hot air and hot steam/air mixtures were used. At the beginning of the experiment the specimen had room temperature; due to heat transfer from the leaking fluid to the specimen and the thermal expansion of the specimen, the crack width was reduced, and the leakage rate significantly decreased. The leakage process including fluid-structure interaction was simulated using the coupling of the thermal-hydraulic system code ATHLET, which has been developed and validated by GRS, and the structure mechanical finite element code ANSYS Mechanical.

Modelling of multidimensional effects in thermal-hydraulic system codes under asymmetric flow conditions – Simulation of ROCOM tests 1.1 and 2.1 with ATHLET 3D-Module

The implementation and validation of multi-dimensional (multi-D) features in thermal-hydraulic system codes aims to extend the application of these codes towards multi-scale simulations. The main goal is the simulation of large-scale three-dimensional effects inside large volumes such as piping or vessel. This novel approach becomes especially relevant during the simulation of accidents with strongly asymmetric flow conditions entailing density gradients. Under such conditions, coolant mixing is a key phenomenon on the eventual variation of the coolant temperature and/or boron concentration at the core inlet and on the extent of a local re-criticality based on the reactivity feedback effects. This approach presents several advantages compared to CFD calculations, mainly concerning the model size and computational efforts. However, the range of applicability and accuracy of the newly implemented physical models at this point is still limited and needs to be further extended. This paper aims at contributing to the validation of the multi-D features of the system code ATHLET based on the simulation of the Tests 1.1 and 2.1, conducted at the test facility ROCOM. Overall, the multi-D features of ATHLET predict reasonably well the evolution from both experiments, despite an observed overprediction of coolant mixing at the vessel during both experiments.

3-D COUPLED SIMULATION OF A VVER 1000 WITH PARCS/ATHLET

A new OECD/NEA benchmark entitled “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster” is starting. This benchmark, based on high quality measurements performed at the NPP Rostov Unit 2, aims to validate and assess high fidelity multi-physics simulation code capabilities. The Benchmark is divided in two phases: assembly wise and pin-by-pin resolution of steady-state and transient multi-physics problems. Multi-physics simulation requires the generation of parametrized few-group cross-sections. This task used to be done with deterministic (2-D) lattice codes, but in the past few years the Monte-Carlo code SERPENT has demonstrate its ability to generate accurate few-group homogenized cross-section without approximations, neither on the geometry nor in the nuclear data. Since the whole core SERPENT models for production of such cross-section libraries would be computationally costly (and the standard 2-D approach may introduce unnecessary large approximations), 3-D models of each assembly type in infinite radial lattice configurations have been created. These cross-sections are then used to evaluate effective multiplication factors for different core configurations with the diffusion code PARCS. The results are compared with the reference SERPENT calculations. In the next step, a thermal-hydraulic model with the system code ATHLET applying an assembly-wise description of the core (i.e. one channel per fuel assembly) has been developed for coupled PARCS/ATHLET transient test calculations. This paper describes in detail the models and techniques used for the generation of the few-group parameterized cross section libraries, the PARCS model and the ATHLET model. Additionally, a simple exercise with coupled code system PARCS/ATHLET is presented and analysed.

Safety cases for design-basis accidents in LWRs featuring passive systems Part 2 – Numerical investigations

This paper deals with the improvement and validation of numerical tools for the simulation of design basis accidents in nuclear power plants equipped with passive safety systems. Numerical models are implemented in the framework of the 1-D thermal–hydraulic system code ATHLET developed by GRS. Experimental reference data for the validation were obtained at the INKA test facility, a model of the KERENA reactor, reproducing the passive safety systems nearly at full scale.The validation effort focuses firstly on the accuracy of the models for the single passive components, and secondly on the ability of the numerical simulation to reproduce the interaction of all components of the KERENA design under realistic conditions as reproduced in the INKA test facility. Thermal-hydraulic models are presented and validated for two passive components of the KERENA reactor: the passive pressure pulse transmitter and the pressure-controlled flooding valve. Finally, the full model of the INKA facility, including these and other passive components, is discussed and numerical results for simulations reproducing three different design basis accidents are validated by comparison with the corresponding experimental data.

New finite element neutron kinetics coupled code system FENNECS/ATHLET for safety assessment of (very) Small and Micro Reactors

Cores of (very) small and medium size reactor–(v)SMR – are often characterized by irregular geometries. Monte Carlo methods are not yet mature enough for transient applications. To perform coupled transient safety assessments of (v)SMRs, the Finite ElemeNt NEutroniCS (FENNECS) code is being developed at GRS. It solves the time-dependent and steady-state three-dimensional few-group diffusion equation in Galerkin finite element representation using upright triangular prisms with linear basis functions as spatial elements. FENNECS is also coupled with the GRS thermal-hydraulic system code ATHLET. For the special meshing, an external meshing tool is being developed as a Python software module. This paper considers the Heat Pipe Micro Reactor core. For the generation of cross sections to be used in FENNECS, Monte Carlo Serpent models are developed for two configurations: all absorber faces of control drums are turned out of the core (All Rods Out – ARO – configuration) and all absorber faces are turned in (All Rods In – ARI – configuration). The Serpent multiplication factors are in good agreement with the benchmark. In the FENNECS model, control drums are approximated by a 12-edges polygon or by a 24-edges polygon. The FENNECS multiplication factors agree well with Serpent reference results for both configurations. This demonstrates the applicability of FENNECS to (v)SMRs.

Advanced modelling of complex boron dilution transients in PWRs – Validation of ATHLET 3D-Module against the experiment ROCOM E2.3

Inherent boron dilution in PWRs constitutes one of the key issues in nuclear safety research. This issue has been widely investigated analytically and experimentally through several test facilities in Germany. Besides the knowledge gained on the matter, the experimental research has contributed to the creation of large databases for the validation of computational tools used in nuclear safety analyses, such as system and computational fluid dynamic (CFD) codes. In this paper, the multidimensional features of the thermal–hydraulic system code ATHLET are assessed based on the results from the boron dilution experiment E2.3, carried out at the ROCOM test facility. The aim of the present paper is twofold; first contributing to the validation of the 3D-Module of ATHLET, as well as assessing the numerical diffusion introduced by the code during the calculation of boron transport and fluid mixing in the developed pseudo multidimensional vessel throughout the progression of the boron dilution event. Overall, the obtained results with ATHLET 3.1A are satisfactory and meet the validation exercise goals, namely in downcomer annulus and lower plenum. However, the limitations of the code to simulate fluid mixing in the cold legs lead to some deviations in the results at the vessel inlet nozzles.

Sub‐channel‐based multiscale thermal‐hydraulic analysis of a medium‐power, lead‐cooled fast reactor

The lead-cooled fast reactor (LFR) offers enhanced safety and reliability with the fine properties of liquid lead and lead alloy. To study accurately the thermal characteristics of fast reactors, the multiscale thermal-hydraulic coupling simulation is an effective way. Multiscale coupling based on the sub-channel code has evident advantages on the analysis of fuel assemblies. In this study, a multiscale thermal-hydraulic analysis of a forced-circulation, medium-power LFR under steady-state and transient conditions is performed with the system code ATHLET and sub-channel code KMC-SUBtraC which was developed based on the previous version by modifying the pressure drop correlations and adding the assembly-level calculation. The codes are one-way-coupled, with good efficiency and precision. Transient verification of the sub-channel code is conducted with the CFD code. In the steady-state analysis of M2LFR-1000, mass flow and temperature distributions of the assemblies, sub-channels, and fuel rods in the hottest assembly are analyzed and the safety performance is investigated. In the transient analysis, two typical DECs (unprotected overpower transient and ULOF+ULOHS) are simulated and the multiscale thermal-hydraulic characteristics are analyzed. With the negative reactivity feedback, the variations of the temperatures of the coolant and fuel rods are within the safe limits, which shows the inherent safety of the reactor. And the results indicate that the loss of primary flow could increase the risk of cladding corrosion.

Advanced Natural Circulation Reduced-Order Model With Inclined Channel for Low Pressure Conditions

Abstract Natural circulation with two-phase flow is a nonlinear dynamical systems, which can show a very complex and strange behavior under specific conditions. The application of stability analysis requires a large computational effort and is cumbersome in case of prediction the dynamical behavior by system codes alone. Therefore, model-order reduction techniques are used to compensate this disadvantage by coupling with a bifurcation code such as MatCont. A reduced-order model is derived by employing the proper orthogonal decomposition (POD) to analyze the stability landscape of a low pressure natural circulation system representing passive safety systems such as the containment cooling condenser. The required full-order model contains a classical natural circulation loop with a heated section and a riser. The two-phase region is modeled by a drift–flux mixture model. The reliability of the full-order model is investigated by comparison with a reference model by the validated system code ATHLET.

External Hazard Coinciding With Small Break LOCA—Thermohydraulic Calculation With System Code ATHLET

Abstract The safety behaviors of a nuclear power plant (NPP) after an external hazard-initiated event, as well as after a small break (SB) loss of coolant accident (LOCA), are already well known as part of the analyses made for standard license application. The coincidence of both events leads to a beyond-design basis consideration. Such a combination of both event categories is investigated by means of the thermohydraulic system code ATHLET. The scenario assumes an external event with a LOCA caused by induced vibrations on a small pipe attached to the primary circuit, although all pipes are designed to withstand the loads created by such an external event. Furthermore, in the context of both robustness and enveloping analyses, both a loss of offsite power (LOOP) and an unavailability of the emergency diesel power supply are postulated. The NPP in the scenario considered only has access to the passive accumulators and to systems supplied via the safeguard emergency diesel engines (second quartet of emergency diesel engines), which are housed in the bunkered emergency feed building. The dedicated type of external event itself is not in focus, but rather the thermohydraulic behavior of the NPP is considered. Apart from the model's assumptions, the accident sequence is explained in detail. The remaining systems for emergency core cooling are capable of handling the LOCA under such demanding boundary conditions. Long-term cooling can be ensured. Furthermore, heat removal out of the core is always sufficient. Eventually, all safety protection objectives have been complied for this beyond-design basis scenario.

ATHLET extensions for the simulation of supercritical carbon dioxide driven power cycles

Abstract The Fukushima accident reveals the need for additional safety systems for nuclear power plants. One promising option is the supercritical carbon-dioxide (sCO2) heat removal system, which consists of a simple Brayton cycle. This study provides an overview of the extensions and validation of the thermal-hydraulic system code ATHLET for the simulation of sCO2 power cycles, especially with regard to the sCO2 heat removal system. The properties of CO2, heat transfer and pressure drop correlations, as well as compact heat exchanger and turbomachinery modelling are considered.

Thermal-hydraulic insights during a main steam line break in a generic PWR KONVOI reactor with ATHLET 3.1A

Abstract The present paper gathers the main insights obtained during the numerical simulation of a 10 % main steam line break (MSLB) in a generic German PWR KONVOI reactor with the thermal-hydraulic system code ATHLET 3.1A. The contents of this paper are focused first on the transient thermal-hydraulic calculation during affected steam generator (SG) 1 boil-off and subsequently on the multidimensional fluid mixing study of the overcooled water stream and the coolant in the reactor pressure vessel. With this aim, the boundary conditions from the test PKL G3.1, carried out at the PKL test facility in the framework of the OECD/PKL-II project, are implemented in the simulation over the plant nominal parameters from the KONVOI reactor. The thermal-hydraulic and fluid mixing results obtained in the simulations are qualitatively assessed against suitable experimental data from the PKL and ROCOM test facilities, showing a good agreement between simulation and test behaviour.

Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies

Abstract The OECD/NEA Benchmark on NPP Kalinin Unit 3 “Switching-off of one of the four operating main circulation pumps at nominal power” based on measurements is at its end phase. A new benchmark is starting based on measurements from the NPP Rostov Unit 2 “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster into the VVER-1000 core”. The Rostov-2 core is loaded with assemblies of a new modern type “TVS-2M” which differs in construction in comparison with the “TVSA”-type implemented in the Kalinin-3 core. The goal of the performed study is to determine the thermohydraulic differences, during steady-state operation, between the two cores based on loads with different assembly types. Results of steady-state simulations are compared for Kalinin-3 and Rostov-2 cores, taken from the respective Benchmarks specifications. To ensure that the observed differences are coming only from the fuel assembly construction, the same axial power density is used for all cases. Calculations are done with an under-development version of the GRS system code ATHLET, which has sub-channel modeling capabilities. Thus, in addition to the typical assembly-wise models, calculations with seven assemblies modeled at the sub-channel (pin-by-pin) level were also performed and analyzed. This work will help the Benchmarks' Rostov-2 participants to set up more accurate core thermohydraulic models.

Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320

Abstract The paper gives a description of conservative analysis of initiating event associated with uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320 of Temelín NPP. This event is included in the group of beyond design basis accidents. The aim of analysis is to determine also the time interval which is necessary for interventions leading to the deterrence of fuel damage. The failure of operator intervention to isolate dilution routes at intervals shorter than 30 min is assumed. Since the response of the whole NPP system influences the course of safety important parameters of the reactor core, the calculations were made by an externally coupled version of the 3D reactor dynamic code DYN3D and the thermohydraulic system code ATHLET. It is shown that, in addition to exceeding the DNBR limit of more than 99 min from the start of the transient, the remaining safety acceptance criteria will not be violated until the end of the calculation.