- Email: [email protected]
COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS
Nuclear Engineering and Design xxx (2016) xxx–xxx
Contents lists available at ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS Javier Jimenez Escalante a,⇑, Valentino Di Marcello a, Victor Sanchez Espinoza a, Yann Perin b a b
Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Germany Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH, Boltzmannstraße 14, 85748 Garching bei München, Germany
h i g h l i g h t s Application of multi-scale thermal-hydraulics approach for the analysis of BWR transients such as the Anticipated Transients Without Scram (ATWS). The multi-scale thermal-hydraulics coupling was implemented inside the SALOME platform using the MEDCoupling interface and some of the NURESIM
platform interpolators. The validation of the coupling was done using the data coming from the OECD/NRC Oskarshamn-2 benchmark.
a r t i c l e
i n f o
Article history: Received 31 May 2016 Received in revised form 26 September 2016 Accepted 2 October 2016 Available online xxxx
a b s t r a c t Validation and qualification of thermal-hydraulics system codes based on measured plant data are essential for their reliability when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, subchannel and system codes. This paper is focused on the application of multi-scale thermal-hydraulics approach (combination of subchannel and system codes) for the analysis of BWR transients such as the Anticipated Transients Without Scram (ATWS). A description of the thermal-hydraulics coupling methodology between ATHLET/COBRA-TF is given. The multi-scale thermal-hydraulics coupling was implemented inside the SALOME platform using the MEDCoupling interface and some of the NURESIM platform interpolators. Finally, the coupled ATHLET/ COBRA-TF model for the Oskarshamn-2 NPP is presented and the main results of the ATWS transient analysis are discussed. The work was carried out in the frame of the European NURESAFE (NUclear REactor SAFEty Simulation Platform) project of the 7th EURATOM Framework Program, aimed at addressing the engineering aspects of nuclear safety of light water reactors. Ó 2016 Elsevier B.V. All rights reserved.
1. Introduction In this paper, the application of a multi-scale thermalhydraulics approach (combination of subchannel and system codes) for the analysis of BWR transients such as the ATWS is described. This work is one of the main contributions of the collaboration between KIT (Karlsruhe Institute of Technology) and GRS (Gesellschaft für Anlagen-und Reaktorsicherheit) to the NURESAFE project. For conducting this work, the coupling implemented by GRS between ATHLET and COBRA-TF within the SALOME platform has been used. One of the objectives of the NURESAFE project (Chanaron et al., 2015) was the development and validation of BWR thermal⇑ Corresponding author. Fax: +49 721 608 23718. E-mail address: [email protected] (J. Jimenez Escalante).
hydraulics (TH) modeling at system scale and subchannel scale. Regarding to the validation of the German system code ATHLET for BWR safety relevant phenomena such as void fraction and critical power using the data available from the international BWR NUPEC BFBT tests, the research work done within the NURESAFE framework by the authors was already published in (Di Marcello et al., 2015). In any transient leading to reactor trip, the shutdown system is assumed to bring the core condition to a safe zero power state with only the core decay heat remaining as a power source. A short description of the ATWS transients is given thereafter. Such an accident could happen if the scram system fails to work during a reactor event (anticipated transient). The types of events considered are those used for designing the plant. According to the US NRC, ATWS are one of the ‘‘worst case” accidents. Hence, any transient leading to a reactor trip combined with a low-probability ATWS condition with a prerequisite of a certain
http://dx.doi.org/10.1016/j.nucengdes.2016.10.001 0029-5493/Ó 2016 Elsevier B.V. All rights reserved.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
2
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
degree of shutdown system failure will thus result in what can be characterized as ‘‘extreme scenarios” (classified as category 5 events). In relation to core reactivity insertion this means that the reactivity effects from Doppler (fuel temperature) and coolant temperature and density (void) will have a predominant influence on the transient evolution. Although operator actions can be expected to mitigate the severity of the final consequences of the transients, in the short term (first instances of the transient), these actions could have a reduce impact due to time scale of the phenomena. In a BWR there are other means to control the core power, though such methods are slower in response than the scram system, e.g. adjusting the MCP (Main Cooling Pump) speed, the feedwater temperature, and injecting boron in the coolant if needed. In addition the negative void reactivity feedback will make the core somewhat less sensitive to reactivity changes with inverse time derivatives of the order of or greater than the fuel rod time constants (about 5 s). Various types of malfunction of the shutdown or scram system can be conceived being of electrical or mechanical nature. In order to mitigate the frequency of occurrence, some measures have already been taken by the reactor design. The Oskarshamn-2 (O2) Swedish nuclear power plant (NPP) experienced a stability event on February 25, 1999 (Kozlowski et al., 2014). A turbine trip took place leading to a loss of feedwater pre-heaters and the control system logic failed resulting in a condition with high feedwater flow and low feedwater temperature without reactor scram. In addition to the initiating event, an interaction of the automatic power and flow control system caused the plant to move into the low flow and high power regime. A combination of the above events culminated in diverging power oscillations, which later on triggered an automatic scram at high power. The OECD/NRC Oskarshamn-2 BWR stability benchmark was established in 2013 (Kozlowski et al., 2013) based on the event. This benchmark provides qualitatively very good data for validating of coupled best estimate thermal-hydraulics and neutron kinetics tools.
2. The ATHLET system code The system code ATHLET Mod 3.0 Cycle B (Analysis of ThermalHydraulics of Leaks and Transient) being developed by GRS is designed to describe the thermal-hydraulics response of the reactor coolant system during normal and off-normal operating conditions (Lerchl et al., 2012). The code has a highly modular structure so that it can include a large spectrum of models and offer a flexible basis for further development. The ATHLET code comprises a thermo-fluid-dynamic module, a heat transfer and heat conduction module, a neutron kinetics module, a general control simulation module, and the general solver of differential equation systems, called FEBE (GS, 2012). The thermo-fluid-dynamic module offers two different systems of equations which are solved in a one-dimensional geometry: (1) a 2-fluid model with 6-equations, with separate conservation equations for liquid and vapor mass, energy, and momentum; (2) 5-equations model, with separate conservation equations for liquid and vapor mass and energy, and a mixture momentum equation. The spatial discretization is performed on the basis of a finitevolume approach. The mass and energy equations are solved within control volumes, and the momentum equations are solved over flow paths connecting the centers of the control volumes. In order to solve the above mentioned equation systems, there are additional parameters which need to be calculated by means of constitutive models or closure equations, and some of them are specifically designed for components (e.g., pumps, valves, steam separator, etc.). The main constitutive models are implemented in ATHLET to describe the fluid properties, wall mass and heat transfer processes, interphase mass and energy transfer between liquid and vapor, velocity differences between liquid and vapor (drift-flux model) and irreversible pressure losses. Besides, ATHLET provides additional constitutive models for the description of sump clogging, interfacial shear forces, mass transfer in the mixture level, diffusion of gasses and critical flow. For the sake of brevity, these models are not described in this paper, but the reader can refer to (GRS, 2012) for a more comprehensive description.
Fig. 1. ATHLET modeling of O2 core.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
2.1. Modeling of the Oskarshamn-2 primary system with ATHLET The BWR Oskarshamn-2 NPP was commissioned in 1975 with a nominal rated power of 1800 MWth. The ATHLET input deck was developed in close collaboration with GRS (Perin, 2013) using the benchmark specifications (Kozlowski et al., 2013, Gajev, 2013). The whole system is composed of a 6-channel core with bypass, an upper and a lower plenum, a recirculation loop with pumps, a downcomer, an ideal separator and the steam domes as well as a steam line. A General Control System (GCSM) is used for the implementation of a steady-state (and transient) control system. In Fig. 1 a schematic representation of the elements is depicted. The first component of the coolant path in ATHLET is the feedwater. It is modeled using a single junction pipe, which is connected to the first cell of the downcomer. The feedwater flow rate is set by a PID-controller comparing the current and desired mixtures level. The downcomer is the path for the liquid water coming from the steam separator and the fresh feedwater flowing
Table 1 ATHLET HFP steady-state results. Parameter
Deviation of ATHLET Mod 3.0 cycle B versus the benchmark data
Steam dome pressure Feedwater temperature Core inlet temperature Total core flow rate Steam flow rate
0.37% 0.05% 0.76% 1.23% 0.60%
3
down to the lower part of the system. The water flows down through the downcomer into the recirculation loop. The plant contains four external recirculation loops with pumps. In this model, only one recirculation loop with one pump is explicitly modeled with a multiplication factor of four. The pump is modeled by its relative differential pressure which is controlled by a unified PID-controller. After the recirculation loop, the water flows into the lower plenum. At the lower plenum exit, the water is distributed to the six core channels and the core bypass. There are overall 444 fuel bundles in the reactor core which can be categorized into six parallel super-channels according to the bundle types and the inlet orifice (i.e. radial position). The core internal and external bypasses are modeled by a single pipe. At the core exit, the water/steam mixture flows into the steam separator. For this model an ideal separator is defined. This means that the carry-under and the carry-over flow quality are respectively 1.0 and 0.0. The saturated steam flows from the ideal separator into the steam dome. There, the flow reaches the top elevation of the whole system. After that the steam flows downwards to the steam line. At the end of the steam line a pressure boundary condition is defined. Using the built-in point kinetics model, an ATHLET stand-alone at Hot Full Power (HFP) steady-state was successfully executed and results were compared against the benchmark data. The core model of ATHLET achieves steady-state conditions after running around 300 s of null transient. A comparison of global values against the benchmark data can be found in Table 1 and shows good agreement between the calculated and measured values. In Fig. 2, the calculated pressure distribution in the whole system at nominal operating conditions is depicted.
Fig. 2. Pressure distribution computed by ATHLET for the O2 HFP steady-state.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
4
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
3. The CTF subchannel code COBRA-TF (Coolant-Boiling in Rod Arrays – Two Fluids) was originally developed by the Pacific Northwest Laboratory in 1980 and had been used and modified by several institutions over the last few decades. COBRA-TF also found use at the North Carolina State University (NCSU) by the Reactor Dynamics and Fuel Management Group (RDFMG) where it has been improved, updated, and subsequently re-branded as CTF (Avramova, 2006, 2007). It solves the transient balance equations based on a separated flow representation of the two-phase flow (9 equations model). It allows LWR (PWR, BWR and VVER) core simulation for best-estimate evaluations of nuclear reactor safety margins at the nodal and sub-channel level. Within the NURESAFE project, it was integrated and coupled inside the SALOME platform for MSLB (Main Steam Line Break) analyses of PWR and VVER, and ATWS studies in BWR. The two-fluid formulation, generally used in thermal-hydraulics codes, separates the conservation equations of mass, energy, and momentum to vapor and liquid. CTF extends this treatment to three fields: vapor, continuous liquid and entrained liquid droplets, which results in a set of nine time-averaged conservation equations. Dividing the liquid phase into two fields is the most convenient and physically reasonable way to handle two-phase flows.
The conservation equations for each of the three fields and for heat transfer from and within the solid structure in contact with the fluid are solved using a semi-implicit, finite-difference numerical technique on an Eulerian mesh. In the Eulerian mesh the time intervals are assumed to be long enough to smooth out the random fluctuations in the multiphase flow, but short enough to preserve any gross flow unsteadiness. The fluid volume is partitioned into a number of computational cells. The equations are solved using a staggered mesh scheme. The phase velocities are obtained at the cell faces, while the state variables – such as pressure, density, enthalpy, and void fraction – are obtained at the cell center. The momentum equations are solved on staggered cells that are centered on the scalar mesh face. The code is able to handle both hot wall and normal flow regimes maps and it is capable of calculating reverse flow, counter flow, and cross flow situations. CTF is developed for using either 3D Cartesian or sub-channel coordinates and, therefore, the code features extremely flexible nodalization for both the thermalhydraulics and the heat-transfer solution. This flexibility allows a fully 3D treatment in geometries amenable to be described in a Cartesian coordinate system and the use of the sub-channel approximation for faster calculations when the flow is principally in one direction.
Fig. 3. O2 Core loading pattern of the 4 different types of fuel assemblies.
Fig. 4. ATHLET-DYN3D-COBRA-TF 3 way coupling.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
5
fuel assembly. The geometrical data provided in the benchmark specifications (Gajev, 2013) were used to build the CTF input deck. In that core loading, there were 4 different types of fuel assemblies, some of them from different vendors, which required a special attention when creating the core description in terms of pressure loss coefficients and channels dimensions. The correspondence between fuel assembly design and channel type is depicted in Fig. 3 following previous publications by Kozlowski (Kozlowski et al., 2014). 4. Coupling between ATHLET/COBRA-TF within SALOME
Fig. 5. ATHLET/COBRA-TF coupling representation.
3.1. Modeling of the Oskarshamn-2 core with COBRA-TF For the Oskarshamn-2 core definition, the 444 fuel assemblies were modeled on a channel wise basis using a single channel per
For the coupling of ATHLET with COBRA-TF, the parallel coupling method is used. The most standard way to implement a TH/TH coupling uses the external coupling method. Several examples can be found in the literature. For instance, the external coupling approach was applied for the coupling of ATHLET with CFD codes (Ansys CFX and OpenFOAM). The parallel coupling is based on domain overlapping between ATHLET/COBRA-TF. The domain overlapping coupling approach used implies that both codes need to model the same region (the reactor core in this case) and only boundary conditions at inlet and outlet of that component have to be transferred. It is a oneway coupling, since no information from the sub-channel code is transferred to the system code. In addition, a three-way coupling between ATHLET/DYN3D/COBRA-TF was developed by GRS within the NURESAFE project. This coupling has been also applied to PWR Main Steam Line Break (Kliem et al., 2016) and BWR Turbine Trip analysis. For the exercise presented in this paper, DYN3D code was not used due to the lack of a proper set of cross section data (in NEMTAB format) for the Oskarshamn-2 core. Thus, the 3 way coupling was simplified into a one way coupling as shown in Fig. 4. With
Fig. 6. Radial power distribution imposed in COBRA-TF.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
6
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
Fig. 7. Radial power distribution imposed in ATHLET.
this simplification, two options were available for running the simulations: either each code uses the power specified in their respective input decks (then it is assured via input deck that the total power is the same) or ATHLET can transfer its power distribution defined in its heat structures into COBRA-TF. The second choice implies that the power imposed in COBRA-TF is coming from a coarser mesh as ATHLET describes the core only with 6 lumped channels and COBRA-TF does it with 444 channels. Both options were tested and the results that can be found in the Section 5.1 of this paper showing that the impact of imposing a coarse 3D power distribution in the void fraction generation is considerable.
In order to initialize the coupling, following the NURESIM platform coupling philosophy, each code involved produces a 2D mesh at core inlet/outlet for managing the data exchange using the REMAPPER interpolation class included in MEDCoupling library (SALOME website, 2016). REMAPPER is in charge of doing the field spatial transposition between different meshes. In Fig. 5 a representation of the actual one way coupling used is depicted. Each COBRA-TF fuel assembly will get its inlet and outlet boundary conditions from the ATHLET code. As for inlet boundary conditions we impose the temperature and mass flow rate. For the outlet boundary conditions, the pressure is imposed.
Fig. 8. Axial power distribution imposed in COBRA-TF and ATHLET.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
For transient simulations, an explicit staggered time step synchronization scheme is used where the time step size used is the smallest one feasible in CTF and ATHLET. This solution increases the computation time but improves the numerical stability. No stability problems have been encountered in any of the coupled simulations. 5. ATHLET/COBRA-TF coupled analysis of a BWR ATWS In order to investigate the thermal-hydraulics coupling developed during the NURESAFE project, a coupled ATHLET/COBRA-TF calculation of the Oskarshamn-2 NPP ATWS event is presented and analysed. The Oskarshamn-2 feedwater transient has been computed using the most accurate 3D power distribution available
7
at the time, which was calculated by a coupled TRACE/PARCS analysis using 444 parallel channels (1 per FA). The power calculation was performed as part of the KIT exercises within the scope of the O2 OECD/NEA benchmark using the TRACE/PARCS coupled system. The radial and axial power profiles at HFP (Hot Full Power) conditions are depicted in Figs. 6 and 8. The total power versus time evolution was taken from the power plant measurements and is represented in Fig. 9. The ATHLET core model, as explained in Section 2.1, is represented by 6 super-channels grouping the different types of fuel assemblies according to the bundle characteristics and the radial position. The radial power distribution imposed in COBRA-TF (see Fig. 6) has been adjusted to the 6 channel core model as shown in Fig. 7 so that it is normalized to 1 in the same way. In order to simulate the feedwater transient under consideration, the following time dependent boundary conditions need to be imposed in the ATHLET model: the feedwater enthalpy and mass flow rate, the turbine pressure and the pump mass flow rate (Fig. 10).
5.1. Selected results of ATHLET/COBRA-TF simulation
Fig. 9. Power versus time boundary condition applied to both COBRA-TF and ATHLET.
In this section a comparison between the results obtained with different versions of the coupled ATHLET/COBRA-TF models will be discussed. In order to test that the multiscale coupling is working correctly and that COBRA-TF is able to reproduce the online time dependent boundary conditions provided by ATHLET, first, a code to code comparison of the main core averaged parameters was conducted. The results are summarized from Figs. 11–13. It is worth mentioning that in these calculations, COBRA-TF always runs in transient mode, meaning that the transient does not start from a converged steady-state solution. That explains why in the first seconds of the transient, COBRA-TF is accommodating its solution to the pseudo-steady state imposed by the boundary conditions (BC) (in fact, during the first 80 s, nothing happens, see Fig. 10).
Fig. 10. Time dependent boundary conditions applied to ATHLET.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
8
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
The core averaged results produced by the ATHLET model (red line) in Fig. 11 are used as a reference to compare against the other two curves. To confirm the validity of the coupling, COBRA-TF should reproduce those values. The green curve in Figs. 11–13 corresponds to the results of COBRA-TF when transferring the power distribution used by ATHLET shown in Fig. 7. In this case, the power is grouped in 6 super-channels and it is transferred from
Fig. 11. Comparison of core averaged Doppler fuel temperature.
ATHLET during the coupling together with the TH core boundary conditions. The blue curve corresponds instead to the 3D power distribution imposed in the input deck of COBRA-TF as described in Fig. 6. The differences between the red and the blue curves are very small during the whole transient. This indicates that the multiscale coupling was properly implemented. In other words, both codes compute the same transient but with different level of details, having in COBRA-TF a one-to-one mapping of fuel assemblies to channels. This model reproduces the core average values predicted by ATHLET and it is more accurate in terms of spatial resolution. When a neutronic code, such as DYN3D, is available, the 3D power distribution can be used to feed both TH models at the same time, as shown in Fig. 4. In Fig. 14 the COBRA-TF computed core coolant density distribution at time 252.7 s is shown. This time point corresponds to the maximum power during the oscillations just before the scram occurs. It can be appreciated that high heterogeneity between neighbouring channels and large gradients between different fuel assemblies are computed. In addition the boiling onset is also strongly dependent on the fuel assembly conditions. In this regards, the COBRA-TF model can be used for safety demonstration analysis. In the same way, ATHLET computes the TH conditions in all the components of the system (see Fig. 15). In both figures the coolant density is displayed using the same scale range. 5.2. Comparison of ATHLET/COBRA-TF results with measured data An exhaustive search in the O2 benchmark measured data was performed in order to identify the most suitable measured data of the ATWS event for the comparison with ATHLET/COBRA-TF results. Unfortunately, some of the measurements refer to parts of the primary system which are not in the ATHLET model, such as turbine speed, position of valves and condenser mass flow rate. Moreover, the pressure at some positions is provided but it is not clear where the measurement was taken, so it cannot be compared to ATHLET results. Suitable signals were found and used for the comparison. Those are: S311K035: total steam flow in the steam line (kg/s) (See Fig. 16). S211K126: reactor pressure (bar) (See Fig. 17).
Fig. 12. Comparison of core averaged coolant density.
Fig. 13. Comparison of core averaged coolant temperature.
Fig. 14. COBRA-TF core coolant density distribution at time 252.7 s of the transient.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
Fig. 16. Comparison of total mass flow rate in the steamline.
S211K404, S211K409, S211K410: Downcomer Level (m) (See Fig. 18). In order to fulfill the requirements of the non-disclosure agreement, Figs. 16–18 contain only normalized values so that no proprietary data can be extracted from them. In general, the computed values by ATHLET match very well the measured data. However, the water level predictions significantly deviates from the measurements as reported in Fig. 18. This discrepancy can be due to the uncertainty in both the measurements
9
Fig. 17. Comparison of the reactor pressure.
and the modeling parameters, which should be further investigated by appropriate sensitivity analyses. As far as the experimental values are concerned, the water level in the core is detected indirectly by a pressure measurement at different elevations and the correct conversion depends on the calibration of the instrumentation. Nevertheless, this parameter might be affected by large uncertainty during accidental transients due to the change of conditions (pressures and temperatures) for which the instrumentation has been designed, as already experienced during the Fukushima accidents (Clayton and Poore, 2013).
Fig. 15. ATHLET coolant density distribution at time 252.0 s of the transient.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
10
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
heaters), thus confirming the validity of the selected coupling approach. Acknowledgments This work has been performed at the Institute for Neutron Physics and Reactor Technology (INR) of the Karlsruhe Institute of Technology (KIT). Also we would like to acknowledge the European Commission as this work has been co-funded within the Euratom FP7 for Nuclear Research and Training Activities (2007–2011) under the contract NURESAFE-323263 (Chanaron, 2012). References
Fig. 18. Comparison of the core water level.
6. Conclusions In this paper, ATHLET and COBRA-TF have been applied to model the O2 BWR NPP using developed standalone ATHLET and COBRA-TF models of the O2 NPP. They were used in coupled thermal-hydraulics simulations within the NURESIM platform. A good agreement in the steady-state at HFP between both codes and the measured data was found. The coupling as well as some details about the multi-scale TH coupling approach has been presented. There is a non-negligible impact on the thermalhydraulics calculations of the feedback parameters, in this case, the coarseness of the power distribution. Thus, having a coarser power distribution in a fine mesh model can produced worse results than those of a coarser model but with consistent assumptions in space and time discretization. The ATHLET/COBRA-TF coupled simulations of the Oskarshamn-2 feedwater event has been presented and discussed. In overall, the comparison of the simulations against some of the measured values shows a very good agreement. Based on the obtained results, it can be concluded that the coupled ATHLET/COBRA-TF code predicts the key thermal-hydraulics phenomena of BWR with fairly good accuracy for the selected ATWS transient (turbine trip followed by loss of feedwater pre-
Avramova, M., 2006. CTF Input Manual. The Pennsylvania State University. Avramova, M., 2007. CTF Theory Manual. The Pennsylvania State University. Chanaron, B., 2012. ‘‘NURESAFE, Nuclear Reactor Safety Simulation Platform – Description of Work”, EU-FP7 project, Grant agreement: 323263. Chanaron, B., Ahnert, C., Crouzet, N., Sanchez, V., Kolev, N., 2015. Advanced multiphysics simulation for reactor safety in the framework of the NURESAFE project. Ann. Nucl. Energy 84, 166–177. http://dx.doi.org/10.1016/j.anucene.2014. 12.013. Clayton, D.A., Poore, W.P., 2013. ‘‘Fukushima Daiichi – A Case Study for BWR Instrumentation and Control Systems Performance during a Severe Accident”. Technical Report, ORNL/TM-2013/154. Di Marcello, V., Jimenez, J., Sanchez, V., 2015. Validation of the thermal-hydraulics system code ATHLET based on selected pressure drop and void fraction BFBT tests. Nucl. Eng. Des. http://dx.doi.org/10.1016/j.nucengdes.2015.04.003. NEDD-14-00862R1. Gajev, I., 2013, D13.11 – Specification and evaluation of the BWR ATWS Benchmark, KTH, NURESAFE internal report. Gesellschaft für Anlagen-und Reaktorsicherheit (GS) mbH, 2012. ATHLET Mod 3.0 Cycle B – Models and Methods, GRS-P-1-Vol 4. Kliem, S., Kozmenkov, Y., Hadek, J., Perin, Y., Fouquet, F., Bernard, F., Sargeni, A., Cuervo, D., Sabater, A., Sanchez-Cervera, S., Garcia-Herranz, N., Zerkak, O., Ferroukhi, H., Mala, P., 2016, Application of the Best-Estimate Plus Uncertainty Approach on a BWR ATWS Transient Using the NURESIM European Code Platform”, submitted to NED. Kozlowski, T., Gajev, I., Lefvert, T., Christer, N., Roshan, S., et al., 2013. BWR Stability Event Benchmark based on Oskarshamn-2 1999 Feedwater Transient. Royal Institute of Technology Stockholm, Sweden, OECD Nuclear Energy Agency. Kozlowski, T., Wysockib, A., Gajevc, I., Xub, Y., Downar, T., Ivanov, K., Magedanz, J., Hardgrove, M., March-Leuba, J., Hudson, N., Ma, W., 2014. Analysis of the OECD/ NRC Oskarshamn-2 BWR stability benchmark. Ann. Nucl. Energy 67, 4–12. http://dx.doi.org/10.1016/j.anucene.2013.09.028. ISSN 0306-4549. Lerchl, G., Austregesilo, H., Schöffel, P., von der Cron, D., Weyermann, F., 2012. ATHLET Mod- 3.0 Cycle B – User’s Manual. Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH. Perin, Y., 2013, D13.21 Description of the ATHLET model for BWR ATWS analysis, NURESAFE deliverable, version 0. SALOME Platform documentation, ‘‘MEDCoupling library”, SALOME website, http://docs.salome-platform.org/latest/gui/MED/medcoupling.html, 2016.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
Contents lists available at ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS Javier Jimenez Escalante a,⇑, Valentino Di Marcello a, Victor Sanchez Espinoza a, Yann Perin b a b
Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen, Germany Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH, Boltzmannstraße 14, 85748 Garching bei München, Germany
h i g h l i g h t s Application of multi-scale thermal-hydraulics approach for the analysis of BWR transients such as the Anticipated Transients Without Scram (ATWS). The multi-scale thermal-hydraulics coupling was implemented inside the SALOME platform using the MEDCoupling interface and some of the NURESIM
platform interpolators. The validation of the coupling was done using the data coming from the OECD/NRC Oskarshamn-2 benchmark.
a r t i c l e
i n f o
Article history: Received 31 May 2016 Received in revised form 26 September 2016 Accepted 2 October 2016 Available online xxxx
a b s t r a c t Validation and qualification of thermal-hydraulics system codes based on measured plant data are essential for their reliability when applied to nuclear power plant analyses. To this purpose, the Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in various validation and qualification activities of different CFD, subchannel and system codes. This paper is focused on the application of multi-scale thermal-hydraulics approach (combination of subchannel and system codes) for the analysis of BWR transients such as the Anticipated Transients Without Scram (ATWS). A description of the thermal-hydraulics coupling methodology between ATHLET/COBRA-TF is given. The multi-scale thermal-hydraulics coupling was implemented inside the SALOME platform using the MEDCoupling interface and some of the NURESIM platform interpolators. Finally, the coupled ATHLET/ COBRA-TF model for the Oskarshamn-2 NPP is presented and the main results of the ATWS transient analysis are discussed. The work was carried out in the frame of the European NURESAFE (NUclear REactor SAFEty Simulation Platform) project of the 7th EURATOM Framework Program, aimed at addressing the engineering aspects of nuclear safety of light water reactors. Ó 2016 Elsevier B.V. All rights reserved.
1. Introduction In this paper, the application of a multi-scale thermalhydraulics approach (combination of subchannel and system codes) for the analysis of BWR transients such as the ATWS is described. This work is one of the main contributions of the collaboration between KIT (Karlsruhe Institute of Technology) and GRS (Gesellschaft für Anlagen-und Reaktorsicherheit) to the NURESAFE project. For conducting this work, the coupling implemented by GRS between ATHLET and COBRA-TF within the SALOME platform has been used. One of the objectives of the NURESAFE project (Chanaron et al., 2015) was the development and validation of BWR thermal⇑ Corresponding author. Fax: +49 721 608 23718. E-mail address: [email protected] (J. Jimenez Escalante).
hydraulics (TH) modeling at system scale and subchannel scale. Regarding to the validation of the German system code ATHLET for BWR safety relevant phenomena such as void fraction and critical power using the data available from the international BWR NUPEC BFBT tests, the research work done within the NURESAFE framework by the authors was already published in (Di Marcello et al., 2015). In any transient leading to reactor trip, the shutdown system is assumed to bring the core condition to a safe zero power state with only the core decay heat remaining as a power source. A short description of the ATWS transients is given thereafter. Such an accident could happen if the scram system fails to work during a reactor event (anticipated transient). The types of events considered are those used for designing the plant. According to the US NRC, ATWS are one of the ‘‘worst case” accidents. Hence, any transient leading to a reactor trip combined with a low-probability ATWS condition with a prerequisite of a certain
http://dx.doi.org/10.1016/j.nucengdes.2016.10.001 0029-5493/Ó 2016 Elsevier B.V. All rights reserved.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
2
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
degree of shutdown system failure will thus result in what can be characterized as ‘‘extreme scenarios” (classified as category 5 events). In relation to core reactivity insertion this means that the reactivity effects from Doppler (fuel temperature) and coolant temperature and density (void) will have a predominant influence on the transient evolution. Although operator actions can be expected to mitigate the severity of the final consequences of the transients, in the short term (first instances of the transient), these actions could have a reduce impact due to time scale of the phenomena. In a BWR there are other means to control the core power, though such methods are slower in response than the scram system, e.g. adjusting the MCP (Main Cooling Pump) speed, the feedwater temperature, and injecting boron in the coolant if needed. In addition the negative void reactivity feedback will make the core somewhat less sensitive to reactivity changes with inverse time derivatives of the order of or greater than the fuel rod time constants (about 5 s). Various types of malfunction of the shutdown or scram system can be conceived being of electrical or mechanical nature. In order to mitigate the frequency of occurrence, some measures have already been taken by the reactor design. The Oskarshamn-2 (O2) Swedish nuclear power plant (NPP) experienced a stability event on February 25, 1999 (Kozlowski et al., 2014). A turbine trip took place leading to a loss of feedwater pre-heaters and the control system logic failed resulting in a condition with high feedwater flow and low feedwater temperature without reactor scram. In addition to the initiating event, an interaction of the automatic power and flow control system caused the plant to move into the low flow and high power regime. A combination of the above events culminated in diverging power oscillations, which later on triggered an automatic scram at high power. The OECD/NRC Oskarshamn-2 BWR stability benchmark was established in 2013 (Kozlowski et al., 2013) based on the event. This benchmark provides qualitatively very good data for validating of coupled best estimate thermal-hydraulics and neutron kinetics tools.
2. The ATHLET system code The system code ATHLET Mod 3.0 Cycle B (Analysis of ThermalHydraulics of Leaks and Transient) being developed by GRS is designed to describe the thermal-hydraulics response of the reactor coolant system during normal and off-normal operating conditions (Lerchl et al., 2012). The code has a highly modular structure so that it can include a large spectrum of models and offer a flexible basis for further development. The ATHLET code comprises a thermo-fluid-dynamic module, a heat transfer and heat conduction module, a neutron kinetics module, a general control simulation module, and the general solver of differential equation systems, called FEBE (GS, 2012). The thermo-fluid-dynamic module offers two different systems of equations which are solved in a one-dimensional geometry: (1) a 2-fluid model with 6-equations, with separate conservation equations for liquid and vapor mass, energy, and momentum; (2) 5-equations model, with separate conservation equations for liquid and vapor mass and energy, and a mixture momentum equation. The spatial discretization is performed on the basis of a finitevolume approach. The mass and energy equations are solved within control volumes, and the momentum equations are solved over flow paths connecting the centers of the control volumes. In order to solve the above mentioned equation systems, there are additional parameters which need to be calculated by means of constitutive models or closure equations, and some of them are specifically designed for components (e.g., pumps, valves, steam separator, etc.). The main constitutive models are implemented in ATHLET to describe the fluid properties, wall mass and heat transfer processes, interphase mass and energy transfer between liquid and vapor, velocity differences between liquid and vapor (drift-flux model) and irreversible pressure losses. Besides, ATHLET provides additional constitutive models for the description of sump clogging, interfacial shear forces, mass transfer in the mixture level, diffusion of gasses and critical flow. For the sake of brevity, these models are not described in this paper, but the reader can refer to (GRS, 2012) for a more comprehensive description.
Fig. 1. ATHLET modeling of O2 core.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
2.1. Modeling of the Oskarshamn-2 primary system with ATHLET The BWR Oskarshamn-2 NPP was commissioned in 1975 with a nominal rated power of 1800 MWth. The ATHLET input deck was developed in close collaboration with GRS (Perin, 2013) using the benchmark specifications (Kozlowski et al., 2013, Gajev, 2013). The whole system is composed of a 6-channel core with bypass, an upper and a lower plenum, a recirculation loop with pumps, a downcomer, an ideal separator and the steam domes as well as a steam line. A General Control System (GCSM) is used for the implementation of a steady-state (and transient) control system. In Fig. 1 a schematic representation of the elements is depicted. The first component of the coolant path in ATHLET is the feedwater. It is modeled using a single junction pipe, which is connected to the first cell of the downcomer. The feedwater flow rate is set by a PID-controller comparing the current and desired mixtures level. The downcomer is the path for the liquid water coming from the steam separator and the fresh feedwater flowing
Table 1 ATHLET HFP steady-state results. Parameter
Deviation of ATHLET Mod 3.0 cycle B versus the benchmark data
Steam dome pressure Feedwater temperature Core inlet temperature Total core flow rate Steam flow rate
0.37% 0.05% 0.76% 1.23% 0.60%
3
down to the lower part of the system. The water flows down through the downcomer into the recirculation loop. The plant contains four external recirculation loops with pumps. In this model, only one recirculation loop with one pump is explicitly modeled with a multiplication factor of four. The pump is modeled by its relative differential pressure which is controlled by a unified PID-controller. After the recirculation loop, the water flows into the lower plenum. At the lower plenum exit, the water is distributed to the six core channels and the core bypass. There are overall 444 fuel bundles in the reactor core which can be categorized into six parallel super-channels according to the bundle types and the inlet orifice (i.e. radial position). The core internal and external bypasses are modeled by a single pipe. At the core exit, the water/steam mixture flows into the steam separator. For this model an ideal separator is defined. This means that the carry-under and the carry-over flow quality are respectively 1.0 and 0.0. The saturated steam flows from the ideal separator into the steam dome. There, the flow reaches the top elevation of the whole system. After that the steam flows downwards to the steam line. At the end of the steam line a pressure boundary condition is defined. Using the built-in point kinetics model, an ATHLET stand-alone at Hot Full Power (HFP) steady-state was successfully executed and results were compared against the benchmark data. The core model of ATHLET achieves steady-state conditions after running around 300 s of null transient. A comparison of global values against the benchmark data can be found in Table 1 and shows good agreement between the calculated and measured values. In Fig. 2, the calculated pressure distribution in the whole system at nominal operating conditions is depicted.
Fig. 2. Pressure distribution computed by ATHLET for the O2 HFP steady-state.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
4
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
3. The CTF subchannel code COBRA-TF (Coolant-Boiling in Rod Arrays – Two Fluids) was originally developed by the Pacific Northwest Laboratory in 1980 and had been used and modified by several institutions over the last few decades. COBRA-TF also found use at the North Carolina State University (NCSU) by the Reactor Dynamics and Fuel Management Group (RDFMG) where it has been improved, updated, and subsequently re-branded as CTF (Avramova, 2006, 2007). It solves the transient balance equations based on a separated flow representation of the two-phase flow (9 equations model). It allows LWR (PWR, BWR and VVER) core simulation for best-estimate evaluations of nuclear reactor safety margins at the nodal and sub-channel level. Within the NURESAFE project, it was integrated and coupled inside the SALOME platform for MSLB (Main Steam Line Break) analyses of PWR and VVER, and ATWS studies in BWR. The two-fluid formulation, generally used in thermal-hydraulics codes, separates the conservation equations of mass, energy, and momentum to vapor and liquid. CTF extends this treatment to three fields: vapor, continuous liquid and entrained liquid droplets, which results in a set of nine time-averaged conservation equations. Dividing the liquid phase into two fields is the most convenient and physically reasonable way to handle two-phase flows.
The conservation equations for each of the three fields and for heat transfer from and within the solid structure in contact with the fluid are solved using a semi-implicit, finite-difference numerical technique on an Eulerian mesh. In the Eulerian mesh the time intervals are assumed to be long enough to smooth out the random fluctuations in the multiphase flow, but short enough to preserve any gross flow unsteadiness. The fluid volume is partitioned into a number of computational cells. The equations are solved using a staggered mesh scheme. The phase velocities are obtained at the cell faces, while the state variables – such as pressure, density, enthalpy, and void fraction – are obtained at the cell center. The momentum equations are solved on staggered cells that are centered on the scalar mesh face. The code is able to handle both hot wall and normal flow regimes maps and it is capable of calculating reverse flow, counter flow, and cross flow situations. CTF is developed for using either 3D Cartesian or sub-channel coordinates and, therefore, the code features extremely flexible nodalization for both the thermalhydraulics and the heat-transfer solution. This flexibility allows a fully 3D treatment in geometries amenable to be described in a Cartesian coordinate system and the use of the sub-channel approximation for faster calculations when the flow is principally in one direction.
Fig. 3. O2 Core loading pattern of the 4 different types of fuel assemblies.
Fig. 4. ATHLET-DYN3D-COBRA-TF 3 way coupling.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
5
fuel assembly. The geometrical data provided in the benchmark specifications (Gajev, 2013) were used to build the CTF input deck. In that core loading, there were 4 different types of fuel assemblies, some of them from different vendors, which required a special attention when creating the core description in terms of pressure loss coefficients and channels dimensions. The correspondence between fuel assembly design and channel type is depicted in Fig. 3 following previous publications by Kozlowski (Kozlowski et al., 2014). 4. Coupling between ATHLET/COBRA-TF within SALOME
Fig. 5. ATHLET/COBRA-TF coupling representation.
3.1. Modeling of the Oskarshamn-2 core with COBRA-TF For the Oskarshamn-2 core definition, the 444 fuel assemblies were modeled on a channel wise basis using a single channel per
For the coupling of ATHLET with COBRA-TF, the parallel coupling method is used. The most standard way to implement a TH/TH coupling uses the external coupling method. Several examples can be found in the literature. For instance, the external coupling approach was applied for the coupling of ATHLET with CFD codes (Ansys CFX and OpenFOAM). The parallel coupling is based on domain overlapping between ATHLET/COBRA-TF. The domain overlapping coupling approach used implies that both codes need to model the same region (the reactor core in this case) and only boundary conditions at inlet and outlet of that component have to be transferred. It is a oneway coupling, since no information from the sub-channel code is transferred to the system code. In addition, a three-way coupling between ATHLET/DYN3D/COBRA-TF was developed by GRS within the NURESAFE project. This coupling has been also applied to PWR Main Steam Line Break (Kliem et al., 2016) and BWR Turbine Trip analysis. For the exercise presented in this paper, DYN3D code was not used due to the lack of a proper set of cross section data (in NEMTAB format) for the Oskarshamn-2 core. Thus, the 3 way coupling was simplified into a one way coupling as shown in Fig. 4. With
Fig. 6. Radial power distribution imposed in COBRA-TF.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
6
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
Fig. 7. Radial power distribution imposed in ATHLET.
this simplification, two options were available for running the simulations: either each code uses the power specified in their respective input decks (then it is assured via input deck that the total power is the same) or ATHLET can transfer its power distribution defined in its heat structures into COBRA-TF. The second choice implies that the power imposed in COBRA-TF is coming from a coarser mesh as ATHLET describes the core only with 6 lumped channels and COBRA-TF does it with 444 channels. Both options were tested and the results that can be found in the Section 5.1 of this paper showing that the impact of imposing a coarse 3D power distribution in the void fraction generation is considerable.
In order to initialize the coupling, following the NURESIM platform coupling philosophy, each code involved produces a 2D mesh at core inlet/outlet for managing the data exchange using the REMAPPER interpolation class included in MEDCoupling library (SALOME website, 2016). REMAPPER is in charge of doing the field spatial transposition between different meshes. In Fig. 5 a representation of the actual one way coupling used is depicted. Each COBRA-TF fuel assembly will get its inlet and outlet boundary conditions from the ATHLET code. As for inlet boundary conditions we impose the temperature and mass flow rate. For the outlet boundary conditions, the pressure is imposed.
Fig. 8. Axial power distribution imposed in COBRA-TF and ATHLET.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
For transient simulations, an explicit staggered time step synchronization scheme is used where the time step size used is the smallest one feasible in CTF and ATHLET. This solution increases the computation time but improves the numerical stability. No stability problems have been encountered in any of the coupled simulations. 5. ATHLET/COBRA-TF coupled analysis of a BWR ATWS In order to investigate the thermal-hydraulics coupling developed during the NURESAFE project, a coupled ATHLET/COBRA-TF calculation of the Oskarshamn-2 NPP ATWS event is presented and analysed. The Oskarshamn-2 feedwater transient has been computed using the most accurate 3D power distribution available
7
at the time, which was calculated by a coupled TRACE/PARCS analysis using 444 parallel channels (1 per FA). The power calculation was performed as part of the KIT exercises within the scope of the O2 OECD/NEA benchmark using the TRACE/PARCS coupled system. The radial and axial power profiles at HFP (Hot Full Power) conditions are depicted in Figs. 6 and 8. The total power versus time evolution was taken from the power plant measurements and is represented in Fig. 9. The ATHLET core model, as explained in Section 2.1, is represented by 6 super-channels grouping the different types of fuel assemblies according to the bundle characteristics and the radial position. The radial power distribution imposed in COBRA-TF (see Fig. 6) has been adjusted to the 6 channel core model as shown in Fig. 7 so that it is normalized to 1 in the same way. In order to simulate the feedwater transient under consideration, the following time dependent boundary conditions need to be imposed in the ATHLET model: the feedwater enthalpy and mass flow rate, the turbine pressure and the pump mass flow rate (Fig. 10).
5.1. Selected results of ATHLET/COBRA-TF simulation
Fig. 9. Power versus time boundary condition applied to both COBRA-TF and ATHLET.
In this section a comparison between the results obtained with different versions of the coupled ATHLET/COBRA-TF models will be discussed. In order to test that the multiscale coupling is working correctly and that COBRA-TF is able to reproduce the online time dependent boundary conditions provided by ATHLET, first, a code to code comparison of the main core averaged parameters was conducted. The results are summarized from Figs. 11–13. It is worth mentioning that in these calculations, COBRA-TF always runs in transient mode, meaning that the transient does not start from a converged steady-state solution. That explains why in the first seconds of the transient, COBRA-TF is accommodating its solution to the pseudo-steady state imposed by the boundary conditions (BC) (in fact, during the first 80 s, nothing happens, see Fig. 10).
Fig. 10. Time dependent boundary conditions applied to ATHLET.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
8
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
The core averaged results produced by the ATHLET model (red line) in Fig. 11 are used as a reference to compare against the other two curves. To confirm the validity of the coupling, COBRA-TF should reproduce those values. The green curve in Figs. 11–13 corresponds to the results of COBRA-TF when transferring the power distribution used by ATHLET shown in Fig. 7. In this case, the power is grouped in 6 super-channels and it is transferred from
Fig. 11. Comparison of core averaged Doppler fuel temperature.
ATHLET during the coupling together with the TH core boundary conditions. The blue curve corresponds instead to the 3D power distribution imposed in the input deck of COBRA-TF as described in Fig. 6. The differences between the red and the blue curves are very small during the whole transient. This indicates that the multiscale coupling was properly implemented. In other words, both codes compute the same transient but with different level of details, having in COBRA-TF a one-to-one mapping of fuel assemblies to channels. This model reproduces the core average values predicted by ATHLET and it is more accurate in terms of spatial resolution. When a neutronic code, such as DYN3D, is available, the 3D power distribution can be used to feed both TH models at the same time, as shown in Fig. 4. In Fig. 14 the COBRA-TF computed core coolant density distribution at time 252.7 s is shown. This time point corresponds to the maximum power during the oscillations just before the scram occurs. It can be appreciated that high heterogeneity between neighbouring channels and large gradients between different fuel assemblies are computed. In addition the boiling onset is also strongly dependent on the fuel assembly conditions. In this regards, the COBRA-TF model can be used for safety demonstration analysis. In the same way, ATHLET computes the TH conditions in all the components of the system (see Fig. 15). In both figures the coolant density is displayed using the same scale range. 5.2. Comparison of ATHLET/COBRA-TF results with measured data An exhaustive search in the O2 benchmark measured data was performed in order to identify the most suitable measured data of the ATWS event for the comparison with ATHLET/COBRA-TF results. Unfortunately, some of the measurements refer to parts of the primary system which are not in the ATHLET model, such as turbine speed, position of valves and condenser mass flow rate. Moreover, the pressure at some positions is provided but it is not clear where the measurement was taken, so it cannot be compared to ATHLET results. Suitable signals were found and used for the comparison. Those are: S311K035: total steam flow in the steam line (kg/s) (See Fig. 16). S211K126: reactor pressure (bar) (See Fig. 17).
Fig. 12. Comparison of core averaged coolant density.
Fig. 13. Comparison of core averaged coolant temperature.
Fig. 14. COBRA-TF core coolant density distribution at time 252.7 s of the transient.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
Fig. 16. Comparison of total mass flow rate in the steamline.
S211K404, S211K409, S211K410: Downcomer Level (m) (See Fig. 18). In order to fulfill the requirements of the non-disclosure agreement, Figs. 16–18 contain only normalized values so that no proprietary data can be extracted from them. In general, the computed values by ATHLET match very well the measured data. However, the water level predictions significantly deviates from the measurements as reported in Fig. 18. This discrepancy can be due to the uncertainty in both the measurements
9
Fig. 17. Comparison of the reactor pressure.
and the modeling parameters, which should be further investigated by appropriate sensitivity analyses. As far as the experimental values are concerned, the water level in the core is detected indirectly by a pressure measurement at different elevations and the correct conversion depends on the calibration of the instrumentation. Nevertheless, this parameter might be affected by large uncertainty during accidental transients due to the change of conditions (pressures and temperatures) for which the instrumentation has been designed, as already experienced during the Fukushima accidents (Clayton and Poore, 2013).
Fig. 15. ATHLET coolant density distribution at time 252.0 s of the transient.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001
10
J. Jimenez Escalante et al. / Nuclear Engineering and Design xxx (2016) xxx–xxx
heaters), thus confirming the validity of the selected coupling approach. Acknowledgments This work has been performed at the Institute for Neutron Physics and Reactor Technology (INR) of the Karlsruhe Institute of Technology (KIT). Also we would like to acknowledge the European Commission as this work has been co-funded within the Euratom FP7 for Nuclear Research and Training Activities (2007–2011) under the contract NURESAFE-323263 (Chanaron, 2012). References
Fig. 18. Comparison of the core water level.
6. Conclusions In this paper, ATHLET and COBRA-TF have been applied to model the O2 BWR NPP using developed standalone ATHLET and COBRA-TF models of the O2 NPP. They were used in coupled thermal-hydraulics simulations within the NURESIM platform. A good agreement in the steady-state at HFP between both codes and the measured data was found. The coupling as well as some details about the multi-scale TH coupling approach has been presented. There is a non-negligible impact on the thermalhydraulics calculations of the feedback parameters, in this case, the coarseness of the power distribution. Thus, having a coarser power distribution in a fine mesh model can produced worse results than those of a coarser model but with consistent assumptions in space and time discretization. The ATHLET/COBRA-TF coupled simulations of the Oskarshamn-2 feedwater event has been presented and discussed. In overall, the comparison of the simulations against some of the measured values shows a very good agreement. Based on the obtained results, it can be concluded that the coupled ATHLET/COBRA-TF code predicts the key thermal-hydraulics phenomena of BWR with fairly good accuracy for the selected ATWS transient (turbine trip followed by loss of feedwater pre-
Avramova, M., 2006. CTF Input Manual. The Pennsylvania State University. Avramova, M., 2007. CTF Theory Manual. The Pennsylvania State University. Chanaron, B., 2012. ‘‘NURESAFE, Nuclear Reactor Safety Simulation Platform – Description of Work”, EU-FP7 project, Grant agreement: 323263. Chanaron, B., Ahnert, C., Crouzet, N., Sanchez, V., Kolev, N., 2015. Advanced multiphysics simulation for reactor safety in the framework of the NURESAFE project. Ann. Nucl. Energy 84, 166–177. http://dx.doi.org/10.1016/j.anucene.2014. 12.013. Clayton, D.A., Poore, W.P., 2013. ‘‘Fukushima Daiichi – A Case Study for BWR Instrumentation and Control Systems Performance during a Severe Accident”. Technical Report, ORNL/TM-2013/154. Di Marcello, V., Jimenez, J., Sanchez, V., 2015. Validation of the thermal-hydraulics system code ATHLET based on selected pressure drop and void fraction BFBT tests. Nucl. Eng. Des. http://dx.doi.org/10.1016/j.nucengdes.2015.04.003. NEDD-14-00862R1. Gajev, I., 2013, D13.11 – Specification and evaluation of the BWR ATWS Benchmark, KTH, NURESAFE internal report. Gesellschaft für Anlagen-und Reaktorsicherheit (GS) mbH, 2012. ATHLET Mod 3.0 Cycle B – Models and Methods, GRS-P-1-Vol 4. Kliem, S., Kozmenkov, Y., Hadek, J., Perin, Y., Fouquet, F., Bernard, F., Sargeni, A., Cuervo, D., Sabater, A., Sanchez-Cervera, S., Garcia-Herranz, N., Zerkak, O., Ferroukhi, H., Mala, P., 2016, Application of the Best-Estimate Plus Uncertainty Approach on a BWR ATWS Transient Using the NURESIM European Code Platform”, submitted to NED. Kozlowski, T., Gajev, I., Lefvert, T., Christer, N., Roshan, S., et al., 2013. BWR Stability Event Benchmark based on Oskarshamn-2 1999 Feedwater Transient. Royal Institute of Technology Stockholm, Sweden, OECD Nuclear Energy Agency. Kozlowski, T., Wysockib, A., Gajevc, I., Xub, Y., Downar, T., Ivanov, K., Magedanz, J., Hardgrove, M., March-Leuba, J., Hudson, N., Ma, W., 2014. Analysis of the OECD/ NRC Oskarshamn-2 BWR stability benchmark. Ann. Nucl. Energy 67, 4–12. http://dx.doi.org/10.1016/j.anucene.2013.09.028. ISSN 0306-4549. Lerchl, G., Austregesilo, H., Schöffel, P., von der Cron, D., Weyermann, F., 2012. ATHLET Mod- 3.0 Cycle B – User’s Manual. Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH. Perin, Y., 2013, D13.21 Description of the ATHLET model for BWR ATWS analysis, NURESAFE deliverable, version 0. SALOME Platform documentation, ‘‘MEDCoupling library”, SALOME website, http://docs.salome-platform.org/latest/gui/MED/medcoupling.html, 2016.
Please cite this article in press as: Jimenez Escalante, J., et al. Application of the ATHLET/COBRA-TF thermal-hydraulics coupled code to the analysis of BWR ATWS. Nucl. Eng. Des. (2016), http://dx.doi.org/10.1016/j.nucengdes.2016.10.001