Volume 81, Issue 2

Over 60 technical experts of the reactor safety research division of the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH are developing and validating reliable methods and computer codes – summarized under the term nuclear simulation chain – for the safety-related assessment for all types of nuclear power plants (NPP) and other nuclear facilities considering the current state of science and technology. This nuclear simulation chain has to be able to simulate and assess all relevant physical processes and phenomena for all operating states and (severe) accidents. In the present contribution, the nuclear simulation chain developed and applied by GRS as well as selected examples of its application are presented. The latter demonstrate impressively the width of its scope and its performance. The GRS codes can be passed on request to other (national as well as international) organizations. This contributes to a worldwide increase of the nuclear safety standards. The code transfer is especially important for developing and emerging countries lacking the financial means and/or the necessary know-how for this purpose. At the end of this contribution, the respective course of action is described.

With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

For decades, the Reactor Simulation and Safety Group at Ruhr-Universität Bochum (RUB) contributes to nuclear safety by computer code validation and model development for nuclear safety analysis. Severe accident analysis codes are relevant tools for the understanding and the development of accident management measures. The accidents in the plants Three Mile Island (USA) in 1979 and Fukushima Daiichi (Japan) in 2011 influenced these research activities significantly due to the observed phenomena, such as molten core concrete interaction and hydrogen combustion. This paper gives a brief outline of recent research activities at RUB in the named fields, contributing to code preparation for plant applications. Simulations of the molten core concrete interaction tests CCI-2 and CCI-3 with ASTEC and the hydrogen combustion test Ix9 with COCOSYS are presented exemplarily. Additionally, the application on plants is demonstrated on chosen results of preliminary Fukushima calculations.

Two kinds of Generation III commercial nuclear power plants have been developed in CNNC (China National Nuclear Corporation), one is a small modular reactor ACP100 having an equivalent electric power 100 MW, and the other is HPR1000 (once named ACP1000) having an equivalent electric power 1 000 MW. Both NPPs widely adopted the design philosophy of advanced passive safety systems and considered the lessons from Fukushima Daichi nuclear accident. As the backbone of the R&D of ACP100 and HPR1000, NPIC (Nuclear power Institute of China) has finished the engineering verification test of main safety systems, including passive residual heat removal experiments, reactor cavity injection experiments, hydrogen combustion experiments, and passive autocatalytic recombiner experiments. Above experimental work conducted in NPIC and further research plan of nuclear safety are introduced in this paper.

In the frame of the nuclear safety research program of the Helmholtz Association HZDR performs fundamental and applied research to assess and to reduce the risks related to the nuclear fuel cycle and the production of electricity in nuclear power plants. One of the research topics focuses on the safety aspects of current and future reactor designs. This includes the development and application of methods for analyses of transients and postulated accidents, covering the whole spectrum from normal operation till severe accident sequences including core degradation. This paper gives an overview of the severe accident research activities at the Reactor Safety Division at the Institute of Resource Ecology.

The THAI (THAI = Thermal hydraulics, Hydrogen, Aerosols, Iodine) experimental programme aims to address open questions concerning the behavior of hydrogen, iodine and aerosols in the containment of water cooled reactors. Since its construction in 2000, THAI programme is being performed in the frame of various national projects (sponsored by German Federal Ministry for Economic Affairs and Energy, BMWi) and two international joint projects (under auspices of OECD/NEA). THAI experimental data have been widely used for the validation and further development of Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes with 3D capabilities. Selected examples of code benchmark exercises performed based on the THAI data include; hydrogen distribution experiment (ISP-47 and OECD/NEA THAI code benchmark), hydrogen combustion behaviour (ISP-49), hydrogen mitigation by PARs (OECD/NEA THAI-2 code benchmark), iodine/surface interactions, iodine mass transfer, and iodine transport and multi-compartment behaviour (EU-SARNET and EU-SARNET2), thermal-hydraulic tests (German CFD-network). In the present paper, a brief overview on the THAI experiments and their role in the containment safety assessment is discussed.

For the safety of spent nuclear fuel assemblies stored in storage pool in the extreme condition where the water is lost completely, a passive spray cooling technique was designed, and its effectiveness has been validated by a functional experiment. The spray cooling characteristics of the spent fuel assembly have also been investigated by the experiment.

This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

For the safety analysis of Supercritical Water-Cooled Reactor (SCWR), one of the challenge tasks is to predict the trans-critical behavior of the reactor system during some accidents. The current safety codes have some shortcomings when the pressure decreases from the supercritical condition to the subcritical state due to the void fraction discontinuity across the critical point. Another challenge is the validation of the system code, which needs the transient experimental data. To overcome the above-mentioned challenges, this paper validates the modified code ATHLET-SC, which is developed based on the pseudo two-phase method. The trans-critical transient data from SWAMUP test facility in Shanghai Jiao Tong University (SJTU) are adopted to compare with the simulation results. The results obtained so far shows that the ATHLET-SC code has good feasibility to the trans-critical simulation of SCWR, and it can be used for transient analysis of SCWR in the future.

While Computational Fluid Dynamics (CFD) is already an accepted industrial tool for single phase flows it is not yet mature for two-phase flows. For this reason the qualification of CFD for reactor safety relevant applications which involve multiphase flows is a present topic of research. At the CFD division of Helmholtz-Zentrum Dresden-Rossendorf (HZDR) hereby beside an application-oriented model development and validation also more generic investigations are done. Thus, the baseline model strategy aims on the consolidation of the CFD-modelling for multiphase to enable reliable predictions for well-defined flow pattern in future. In addition the recently developed GENTOP-concept broadens the range of applicability of CFD. Different flow morphologies including transitions between them can be considered in frame of this concept.

The article provides an overview on the code DYN3D which is a three-dimensional core model for steady-state, dynamic and depletion calculations in reactor cores with quadratic or hexagonal fuel assembly geometry being developed by the Helmholtz-Zentrum Dresden-Rossendorf for more than 20 years. The current paper gives an overview on the basic DYN3D models and the available code couplings. The verification and validation status is shortly outlined. The paper concludes with the current developments of the DYN3D code. For more detailed information the reader is referred to the publications cited in the corresponding chapters.

The discharge could be non-choking or choking, depending on the break shape, length and conditions. This presents a challenge in the calculation of standard problems. A stable experiment of water was performed to study the break flow rate in nozzles of diameter of 1.41 and 2.0 mm with rounded-edge and sharp-edge. The pressure covered the ranges of 0.5 to 29.5 MPa, inlet quality 0 to 1.0 and subcooling up to 350 °C. The results exhibited a close relation of thermal non-equilibrium with pressure. For supercritical pressure a modified equilibrium model in combination with the Bernoulli equation is presented.

The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I&C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I&C architecture and protection zones in the plant IT environment.

In this paper recent developments of various techniques for single-phase and two-phase flow measurements with relevance to liquid metal cooled systems will be presented. Further, the status of the DRESDYN platform for large-scale experiments with liquid sodium is sketched.

This paper reports the design and safety analysis results of the helium cooled solid breeder blanket of the Chinese Fusion Engineering Test Reactor (CFETR). Materials selection and basic structure of the blanket have been presented. Performance analysis including neutronics analysis and thermo-mechanical analysis has shown good results. And the safety analysis of the blanket under Loss Of Coolant Accident (LOCA) conditions has been described. Results showed the current design can deal well with the selected accident scenarios.

The German HTGR fuel development program for the HTR-Modul concept has resulted in a reference design based on LEU UO 2 TRISO coated particle fuel in a spherical fuel element. The coated particles consist of minute uranium particle kernels coated with layers of carbon and silicon carbide. Analyses on quality of as-manufactured fuel, its behavior under HTR-Modul relevant operating and accident conditions have demonstrated excellent performance. Coated particles can withstand high internal gas pressure without releasing their fission products to the environment. International efforts are on-going for further improvement of coated particle fuel to meet the needs of future generation-IV HTR concepts.

There is a large potential for nuclear energy also in the non-electric heat market. Many industrial sectors have a high demand for process heat and steam at various levels of temperature and pressure to be provided for desalination of seawater, district heating, or chemical processes. The future generation of nuclear plants will be capable to enter the wide field of cogeneration of heat and power (CHP), to reduce waste heat and to increase efficiency. This requires an adjustment to multiple needs of the customers in terms of size and application. All Generation-IV concepts proposed are designed for coolant outlet temperatures above 500 °C, which allow applications in the low and medium temperature range. A VHTR would even be able to cover the whole temperature range up to approx. 1 000 °C.