Volume 88, Issue 2

The 33rd German CFD Network of Competence Meeting was held in March 2022 at the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH in Garching, Germany. In 2022 the meeting celebrates its 20th anniversary with 17 scientific presentations, distributed in two main sessions: “Simulation of Reactor Cooling Circuit Flows” and “Simulation of Reactor Containment Flows”. This paper gives an overview of the different contributions, presented at this anniversary meeting, and also provides information on the background and the objectives of the German CFD Network of Competence.

Open-source environments such as the Computational Fluid Dynamics software OpenFOAM are very appealing for research groups since they allow for an efficient prototyping of new models or concepts. However, for downstream developments to be sustainable, i.e. reproducible and reusable in the long term, a significant amount of maintenance work must be accounted for. To allow for growth and extensibility, the maintenance work should be underpinned by a high degree of automation for repetitive tasks such as build tests, code deployment and validation runs, in order to keep the focus on scientific work. Here, an information technology environment is presented that aids the centralized maintenance of addon code and setup files with relation to reactor coolant system safety research. It fosters collaborative developments and review processes. State-of-the-art tools for managing software developments are adapted to meet the requirements of OpenFOAM. A flexible approach for upgrading the underlying installation is proposed, based on snapshots of the OpenFOAM development line rather than yearly version releases, to make new functionality available when needed by associated research projects. The process of upgrading within so-called sprint cycles is accompanied by several checks to ensure compatibility of downstream code and simulation setups. Furthermore, the foundation for building a validation data base from contributed simulation setups is laid, creating a basis for continuous quality assurance.

Spacer grids with mixing vanes have complex geometry and are used to support the fuel rods in nuclear fuel assemblies as well as to improve heat transfer by generating turbulence downstream of the spacer grid in the cores of the pressurized water reactors. To validate the CFD code OpenFOAM for relevant spacer grid geometries, numerical analyses on the OECD/NEA MATiS-H benchmark were performed. The flow behind two different types of spacer grid designs was analysed: split- and swirl-type. Initially, an appropriate inlet velocity profile was generated. In a next step, Computer-Aided Design models of the spacer grids were prepared and then meshed using ANSYS Mesher 19.2. Transient URANS simulations were performed with the k - ω- SST turbulence model and the results were compared with data. Good agreement was obtained for the mean velocity profile and the vorticity in the swirl-type configuration, while the numerical results slightly overestimated the transverse velocity profile at some measurements locations of the split-type configuration. The content of this article was initially presented at the 33rd German CFD Network of Competence Meeting, held on March 22–23 at GRS in Garching, Germany.

Flow-induced vibrations in nuclear power plants may lead to material fatigue, fretting wear, and eventually to loss of component integrity. The consequences might be substantial costs due to long unplanned outages or a fault that requires safety provisions to perform as intended. To avoid these, Fluid-Structure Interaction analyses are performed to understand and predict the complex thermal-hydraulic and structural mechanics phenomena. To further advance the knowledge of solving FSI problems with the help of numerical tools, in the beginning of 2020, the joint industry VIKING project was established in Europe. Further, OECD/NEA initiated in 2021 an FSI Benchmark on FIV that should be finished by the end of 2022 and the final synthesis report should be published in 2023. This paper provides a short overview of the GRS contributions within these two international activities on the prediction of FIV in nuclear power reactors. The content of this article was initially presented at the 33rd German CFD Network of Competence Meeting, held on March 22–23 at GRS in Garching, Germany.

This work is focused on the development and validation of models and methods for the simulation of wall boiling in nuclear engineering applications with the computational fluid dynamics (CFD) code OpenFOAM. The new chtMultiRegionReactingTwoPhaseEulerFoam solver was developed based on the reactingTwoPhaseEulerFoam solver of OpenFOAM Foundation version 7. The solver is used for the simulation of two-phase flow under consideration of wall boiling and conjugate heat transfer (CHT) between solid structure and two-phase fluid regions. The Euler–Euler approach for two-phase flows was used. The heat flux during wall boiling was calculated with the help of the extended Rensselaer Polytechnic Institute wall heat flux partitioning model, in which the convective heat flux between solid wall and two-phase flow with high void fractions was also considered. The solver was validated against experimental data from the OECD/NEA PWR Subchannel and Bundle Tests benchmark. This Nuclear Power Energy Corporation (NUPEC) database provides data for different fuel assembly subchannel geometries at different thermal-hydraulic conditions. 10 experimental runs with different boundary conditions of the benchmark exercise I-1 were simulated with the chtMultiRegionReactingTwoPhaseEulerFoam solver. The solver showed good numerical stability in all examined cases, which captured different boiling regimes with up to cross-section averaged void fractions of 0.6. The results were compared with measured data for the averaged over the cross-section of the investigated geometry void fractions. Good agreement with experimental data was observed.

A CANDU reactor core comprises several hundred horizontal fuel channels spanning a calandria vessel. Loss of sufficient cooling during a severe accident could result in collapse of the core to the bottom of the calandria. A simple computational tool for simulating, in two dimensions, the resulting build-up of a terminal debris bed is described. The tool is used to model a variety of core collapse scenarios. Computed debris beds are generally lower in the middle, ∼10 fuel channels deep, and have higher decay power in their interiors. The initial debris bed porosity is estimated to be 0.65 ± 0.15. High porosity could augment in-vessel hydrogen generation and fission product release during subsequent debris bed heat-up.

Remaining Useful Life (RUL) estimation has been extensively explored in recent years. RUL could be used in deciding the maintenance timeline or inspection interval for the Reaktor Serba Guna – G. A. Siwabessy (RSG-GAS reactor). RSG-GAS reactor is a pool-type research reactor (built by the Interatom Internationale of Germany) and has been operating for more than 30 years to date. This study aimed to propose a Weibull model to find the RUL estimation value of the distribution parameters of the mean time to failure (MTTF). Therefore, the RSG-GAS reactor would be higher safety, longer lifetime and higher reliability with a smaller failure rate including for the PA01-AP01 secondary pump. The research methodology is processing data collection and estimating the parameters of the Weibull model to determine maintenance timeline or inspection intervals based on the MTTF value in case the reliability has reached the targeted percentage. Results show that the RUL estimation has been obtained for the RSG-GAS reactor. In the implemented study, a maintenance timeline has been stipulated for the PA01-AP01 secondary pump (with the model of KSB and type of CPK-S350-400) for the reliability of 90% and RUL estimation of circa 29 days.

Critical (choked) flow is a highly concerning phenomenon in safety analysis for nuclear energy. The discharge mass flow rate prediction is crucial for engineering design and emergency response in case of nuclear accidents. Unfortunately, the critical flow is difficult to predict especially when the two-phase flow exists. The accuracy is based on a deeper understanding of the complex phenomenon of critical flow. The influence of virtual mass force on the two-phase critical flow was seldom concentrated on owing to the lack of suitable critical flow models for studies in detail. This study is based on a developed 6-equation two-phase critical flow model. It is confirmed that the virtual mass force contributes to the stability and convergence of the critical flow simulation and it will impact not only the critical mass flux but also the thermal hydraulic parameters along the discharge duct. The magnitude depends on the geometry of the discharge duct and the upstream condition. It is larger when the duct is longer and the pressure is lower. Furthermore, the virtual mass force for each flow regime was studied in detail with a sensitivity study. The results show that the most sensible condition for the virtual mass force is annular flow along a long tube under relatively low pressure. The future work is to develop a correlation of virtual mass force for critical flow specifically since the correlations in the literature were developed under general two-phase flow process conditions.

The Unit 1 of Kuosheng nuclear power plant is in the decommissioning transition phase and still has spent fuel in the core of reactor now. The time to reach the TAF (top of active fuel) for water level and reach 600/800 °C for cladding temperature are the key parameters in the safety analysis, these affect that the plant has how much time to handle transients. Therefore, to study the water level and cladding temperature for station blackout (SBO) and loss of coolant accident (LOCA) transients in the decommissioning transition phase, the analysis model of Kuosheng nuclear power plant was established by using TRACE code. To evaluate the effect of the decay heat of spent fuel on the water level and cladding temperature, the sensitivity analysis of decay heat was also performed in this study.

To address the issue identified in USNRC’s Generic Letter (GL) 2003-01 that the unfiltered air in-leakage rate through plant’s control room envelope during design basis accident may exceed that assumed in the licensing analysis and thus threat the control room habitability, the control room envelope unfiltered in-leakage Tracer Gas Test (TGT) for Maanshan nuclear power plant (NPP) has been performed in 2020. For future applications, an improved control room dose analysis approach using Alternative Source Terms (AST) has been developed in present study to check whether the TGT results meet regulatory limits and have sufficient safety margins. The AST method follows Regulatory Guide 1.183 (RG 1.183) and the TGT results must be fulfilled the total effective dose criteria set forth in 10 CFR 50.67. Based on the AST approach and the TGT results, the control room personnel dose for Maanshan NPP during Loss of Coolant Accident (LOCA) is 14.35 mSv, and the safety margin is 71.3%. It is sufficient to cover the effects of structural ageing and changes in meteorological data during the control room habitability reassessment and the analysis uncertainty.

In this study we have explored 6 Li +  7 Li fusion evaporation reactions cross sections dependencies on both nuclear level density and various spin combination effects. The reaction cross section was calculated in the energy range of 0.1–16 MeV projectile of 6 Li on the fixed target of 7 Li. The excited compound nucleus ( 13 C) can decay into various channels, and its decay rate in any given channel is proportional to the available phase space, i.e., the corresponding level density of it which is explained in the present study. In the present study, LISE ++ , PACE4, NRV and GEMINI codes were used to determine cross section of evaporation residues cross sections of 13 C.

Accelerator Driven sub-critical System (ADS), which employs the high-energy proton beam generated by accelerator to bombard the target nucleus and generate spallation neutrons as external neutrons to drive and maintain the operation of its sub-critical reactor, is of great significance in nuclear waste treatment and disposal. As the instability of proton beam would affect the power level of the reactor and threaten the safety of ADS, Beam Trip (BT) and Beam OverPower (BOP) are commonly considered to be its two typical transient accidents. As for the sub-critical reactor, the Transient OverPower (TOP) is also one of typical transient accidents that should be considered, which is mainly caused by reactivity insertion under certain cases, such as SGTR (Steam Generator Tube Rupture) accident. For the subcritical reactors, the transient evolution behaviors are strongly affected by the subcriticality value. On the one hand, the subcriticality values of ADS design should take safety margin and power gain into consideration. On the other hand, the subcriticality value is variable with the burnup of reactors. So it is necessary to study the safety characteristics of the subcritical reactors under different subcriticality values, in this paper, the transient safety characteristics of a single channel for XADS under BT, BOP and TOP accidents of different subcriticality values were investigated by using MPC-LBE code.