Volume 84, Issue 5

The transition from Generation 2 to Generation 3/3+ and 4 reactors, as well as the development of small modular reactors (SMR), place new demands on computational programs designed to simulate conditions of normal operation, operational occurrences, design basis accidents and severe accidents. On the one hand, most passive safety systems of advanced and innovative plants operate at low pressures even down to vacuum conditions and the driving forces are low compared to active systems. On the other hand, the containment is no longer just a barrier to retain radioactive material in the event of leakage of the cooling system, but it is an important link in the passive cooling chain. This requires an expansion and improvement of the existing simulation programs for the cooling circuit and containment, as well as the realization of a coupling between these simulation programs. The new AC 2 program package combines the proven simulation codes ATHLET/ATHLET-CD and COCOSYS in one software suite to hit this target. The individual components of the suite are continuously extended and validated for their application to novel safety systems. This makes it possible to simulate the entire spectrum of accidents for Generation 3/3+, 4 and light water cooled SMR systems with just one program package. This publication gives an overview of the current state of development of AC 2 and its individual modules.

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

The heat transfer coefficient is an essential measure in the predesign of supercritical water-cooled reactors (SCWRs). At supercritical pressures, three distinct heat transfer modes exist: normal, improved, and deteriorated. The heat transfer behavior of supercritical water in the pseudo-critical range is different from that of single-phase fluids in the subcritical range. These heat transfer modes differ from those of single-phase flow at subcritical pressures, resulting in an unusual behavior of the heat transfer coefficients. Moreover, during accidental scenarios, when the operating pressure is reduced from supercritical to subcritical conditions, a boiling crisis may occur. During pressure reduction, temporary phenomena such as superheating of the cladding temperature can endanger the safe operation of SCWRs. In order to analyze operational and accidental scenarios of SCWRs, thermal-hydraulic system codes such as ATHLET are applied. However, the prediction capabilities of thermal-hydraulic system codes rely on a comprehensive validation work based on experimental data. This study presents an extensive analysis of the applicability of ATHLET at the near-critical pressure range. ATHLET is assessed against the LESHP-database and two trans-critical transient experiments. At supercritical pressures, the heat transfer coefficient correlations are evaluated with regard to their prediction accuracy and numerical problems including the “multiple solutions problems”. The trans-critical transient experiments are used to test the prediction capability of ATHLET with respect to transient heat transfer phenomena including critical heat flux, film boiling and return to nucleate boiling.

The Fukushima accident reveals the need for additional safety systems for nuclear power plants. One promising option is the supercritical carbon-dioxide (sCO 2 ) heat removal system, which consists of a simple Brayton cycle. This study provides an overview of the extensions and validation of the thermal-hydraulic system code ATHLET for the simulation of sCO 2 power cycles, especially with regard to the sCO 2 heat removal system. The properties of CO 2 , heat transfer and pressure drop correlations, as well as compact heat exchanger and turbomachinery modelling are considered.

Verification and validation are basic quality assurance elements in code development and essential for code release. Therefore, the codes of AC 2 (ATHLET – ATHLET-CD – COCOSYS) are tested on separate effect tests, integral tests as well as plant scenarios to verify and validate the models after new implementation or updates. The verification assures that the models are implemented and working correctly while the validation checks if the models predict the right phenomena and combined with other models and modules. The selected experiments are summarized in GRS's validation matrices, which in turn are based on the CSNI validation matrices derived from OECD/WGAMA task groups as well as current activities on experimental test campaigns. For ATHLET several test series are used to cover a wide range of phenomena which can occur in PWR, BWR and VVER. Additionally, plant transients are considered for German LWR. The ATHLET-CD validation matrix contains experiments covering most phenomena which can occur during a severe accident. But due to the interaction of several effects even in small scale experiments mainly integral experimental campaigns are used for the validation. Over the last decades the validation of the AC 2 codeds ATHLET and ATHLET-CD has reached a high degree of fulfilment of GRS's validation matrices over all code versions. Innovative and advanced reactor concepts come with new or newly relevant phenomena, which AC 2 needs to provide models for. Extending the validation base of AC 2 for these models is one challenge for further code validation efforts besides the on-going update of the validation basis to recent code versions.

The stabilization of molten core material in the lower head in case of a severe accident by external cooling of the reactor vessel is regarded as an effective severe accident management measure. In the experiments LIVE-L10 and -L11 the late phase melt pool behaviour of the corium is investigated under different cooling conditions – the former under sub-cooled convection, the latter under nucleate boiling conditions. In this work the experiments are calculated with the severe accident analysis code AC 2 – ATHLET-CD 3.1A. Objective of the simulations conducted is the analysis and assessment of the code's capability to simulate the most relevant phenomena that occur during the tests. The simulations are performed with two different lower head modules implemented in ATHLET-CD, AIDA ( A nalysis of the I nteraction between Core D ebris and the reactor pressure vessel during severe A ccidents) and LHEAD (extended L ower Head module). The simulation results, analysed in comparison with the experimental results, show the capability of both modules to reproduce the respective experiments.

The GRS program package AC 2 with its codes ATHLET/ATHLET-CD and COCOSYS aims for the reliable computational simulation of significant phenomena occurring during normal operation, design basis accidents, and severe accidents in the cooling circuit and containment of a nuclear power plant. To keep the modelling at the state-of-the-art, continuous development and validation is required. This is accomplished through participation in several national and international experimental research programs, where AC 2 or one of its codes are assessed against both separate effect tests and integral tests. This paper exemplifies the status of validation and application of COCOSYS by means of calculations of iodine chemistry and molten corium/concrete interaction after reactor pressure vessel rupture. Further, calculations using the external 3D module CoPool coupled to COCOSYS on thermal stratification in large water pools are discussed. The examples given demonstrate the progress of the COCOSYS development and the capability to simulate phenomena in the containment during incidents and accidents with good results. Future applications comprise the entire spectrum of incidents and accidents for Generation III/III+ systems with just one program package.

Nuclear power plant containment is the last barrier, as stated in the worldwide accepted in-depth safety principle, between a nuclear reactor and the environment. During various accident scenarios, fission products can exit the fuel and primary circuit, entraining the containment and later the environment. One of the most important fission products exiting the corrupted fuel pins is iodine. Iodine is a strongly bioactive chemical species with a very complex in-containment physical-chemical behavior. The fission product transport inside the containment as well as the release into the environment takes an important role in the preliminary safety analysis report (PSAR). Analyses presented in PSAR are calculated by numerous computational codes. The validity of these results should be checked. The only chance to do so is to validate the computational tool on validation experiments, which represent single effect behavior as well as a complex behavior on integral tests. Validation experiments presented in this paper are from an international Behavior of Iodine Project (BIP) which was conducted under the Organization for Economic Co-operation and Development – Nuclear Energy Agency (OEC-NEA). The single effect tests are illustrated by four representative experiments, namely the G-01 studying the iodine deposition and resuspension on steel coupon in gas phase, the G-04 studying the iodine deposition and resuspension on painted surface in gas phase, the G-06 studying the humidity effect on iodine deposition on painted coupons in gas phase and finally the AECL-2 studying the iodine deposition on painted coupons in water phase. The integral test is represented by the RTF P9T1, where the complex iodine chemistry including the response to pH change is studied. The scope of this paper is to reveal whether the COCOSYS code is capable to bring satisfactory results of a complex in-containment iodine behavior with recommended and basic setup of computational models and parameters.

In case of a postulated loss-of-coolant accident (LOCA) in a light water reactor, core degradation might occur if emergency measures fail. During this process a large amount of hydrogen might be generated by the oxidation of the fuel rod cladding. If the hydrogen leaks into the containment of a pressurized water reactor and if passive autocatalytic recombining fails, it might form a combustible mixture with air. In case of ignition, pressure peaks can occur that are relevant for containment integrity. During LOCA the third relevant component of the containment atmosphere is steam. Test TH-29.3 aims at expanding knowledge regarding the interaction of steam and light gas (e.g. hydrogen) and was conducted in the THAI test facility. During the test steam and helium (as ta substitute for hydrogen) are injected into the test vessel. A steam-helium cloud expands downwards and a helium-rich layer forms at the boundary of the cloud. The test is simulated by the use of the lumped parameter code COCOSYS (Containment Code System) which itself is part of the software package AC 2 developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH.

The paper presents analysis results of melt spreading and core-concrete interactions in the containment of a WWER-1000 plant during the ex-vessel phase of a severe accident. The failure of the vessel takes place 8 h 35 min after the initiation of the accident. It has been assumed that the whole area of the containment floor is available for spreading, i.e. the door between the reactor cavity and the main part of the containment is not locked. The melt flow rate from the reactor pressure vessel was used as a boundary condition. The simulation of the melt spreading was performed with the LAVA code. The calculated spreading area varies from 60 to 100 m 2 depending on the assumed values of the melt properties. The results from the LAVA calculations were used in parallel for COCOSYS and MELCOR calculations to study the core-concrete interactions. From the analyses it turned out: a larger spreading area leads to a faster cooling of the melt in the initial period of the accident, but in the long term the temperatures are the same. 60 h after start of the ex-vessel phase, the melt is not stabilised.

Countercurrent gaseous flows may occur under certain conditions e.g. in narrow flow paths between compartments within containments of nuclear power plants (NPP) during design basis and beyond design basis accidents. Their potential effect on thermal-hydraulics within the containment is not sufficiently investigated yet. This publication focuses on modelling and numerical investigation of the physical phenomenon of pressure loss due to turbulence in such countercurrent gas-gas flows. Scenarios observed in the test facilities PANDA and THAI+ are being investigated. The important regimes of the countercurrent layered gaseous flows are identified and a mathematical model incorporating all the required regimes is introduced, implemented in the GRS program COCOSYS, tested and discussed in detail. The effects of the new model are numerically investigated. By means of a comparison with a CFD simulation, the model parameters are evaluated. It is suggested to improve the accuracy of the model with the help of future experiments.