Assessment of CTF boiling transition and critical heat flux modeling capabilities using the OECD/NRC BFBT and PSBT benchmark databases

The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The modeling needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermalhydraulics modeling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan. From 1987 to 1995, NUPEC performed steady-state and transient critical power and departure from nucleate boiling (DNB) test series based on the equivalent full-size mock-ups. Considering the reliability not only of the measured data, but also other relevant parameters such as the system pressure, inlet sub-cooling and rod surface temperature, these test series supplied the first substantial database for the development of truly mechanistic and consistent models for boiling transition and critical heat flux. Over the last few years the Pennsylvania State University (PSU) under the sponsorship of the U.S. Nuclear Regulatory Commission (NRC) has prepared, organized, conducted and summarized the OECD/NRC Full-size Fine-mesh Bundle Tests (BFBT) Benchmark. The international benchmark activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and Japan Nuclear Energy Safety (JNES) organization, Japan. Consequently, the JNES has made available the Boiling Water Reactor (BWR) NUPEC database for the purposes of the benchmark. Based on the success of the OECD/NRC BFBT benchmark the JNES has decided to release also the data based on the NUPEC Pressurized Water Reactor (PWR) subchannel and bundle tests for another follow-up international benchmark entitled OECD/NRC PWR Subchannel and Bundle Tests (PSBT) benchmark. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known subchannel code COBRA-TF, namely CTF, to the critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmark

Dynamic Behaviour of a Continuous Heat Exchanger/Reactor after Flow Failure

The intensified technologies offer new prospects for the development of hazardous chemical syntheses in safer conditions: the idea is to reduce the reaction volume by increasing the thermal performances and preferring the continuous mode to the batch one. In particular, the Open Plate Reactor (OPR) type “reactor/ exchanger” also including a modular block structure, matches these characteristics perfectly. The aim of this paper is to study the OPR behaviour during a normal operation, that is to say, after a stoppage of the circulation of the cooling fluid. So, an experiment was carried out, taking the oxidation of sodium thiosulfate with hydrogen peroxide as an example. The results obtained, in particular with regard to the evolution of the temperature profiles of the reaction medium as a function of time along the apparatus, are compared with those predicted by a dynamic simulator of the OPR. So, the average heat transfer coefficient regarding the “utility” fluid is evaluated in conductive and natural convection modes, and then integrated in the simulator. The conclusion of this study is that, during a cooling failure, a heat transfer by natural convection would be added to the conduction, which contributes to the intrinsically safer character of the apparatus

Quench front progression in a superheated porous medium: experimental analysis and model development

In case of severe accident in a nuclear reactor, the fuel rods may be highly damaged and oxidized and finally collapse to form a debris bed. Removal of decay heat from a debris bed is a challenging issue because of the difficulty for water to flow inside. Currently, IRSN has started experimental program PEARL with two experimental facilities PRELUDE and PEARL, to investigate the reflood process at high temperature, for various particle sizes. On the basis of PRELUDE experimental results, the thermal hydraulic features of the quench front have been analysed and the intensity of heat transfers was estimated. From a selection of experimental results, a reflooding model was improved and validated. The model is implemented in the code ICARE-CATHARE developed by IRSN which is used for severe accident reactor analysis

Heat exchanger/reactors (HEX reactors): Concepts, technologies: State-of-the-art

Process intensification is a chemical engineering field which has truly emerged in the past few years and is currently rapidly growing. It consists in looking for safer operating conditions, lower waste in terms of costs and energy and higher productivity; and away to reach such objectives is to develop multifunctional devices such as heat exchanger/reactors for instance. This review is focused on the latter and makes a point on heat exchanger/reactors. After a brief presentation of requirements due to transposition from batch to continuous apparatuses, heat exchangers/reactors at industrial or pilot scales and their applications are described

Introduction to Nuclear Propulsion: Lecture 15 - Nuclear Test Operations

The test operation of nuclear power plants, specifically nuclear rockets, bears some interesting similarities to the operation of chemical rocket tests as well as, of course, many differences. A significant feature common to both nuclear and chemical rocket tests is that all the fuel for the entire operation is loaded at the start of the test. As a direct consequence of this fact, the operation of nuclear power plants must be surrounded with adequate safety precautions, as is indeed the case in the operation of chemical rockets, A second direct consequence is that in both types of testing a very thorough and complete checkout is made before starting the test

Design and Performance of Hybrid Control Rod For Passive IN-core Cooling System

Department of Nuclear EngineeringProtection of the public and the environment from undue radiation hazards is a definition of nuclear safety. Although there are various safety systems in nuclear power plants to achieve the nuclear safety, Fukushima-Daiichi accident showed the vulnerabilities of the installed safety systems. After the Fukushima accident, various passive safety systems and strategies are under development to cope with the postulated accidents. The majority of passive safety systems concentrated to inject emergency core coolant (ECC) or feedwater with the circuits comprise many pipelines and valves. In station blackout condition, the pressure of reactor vessel would be higher than the ECC injection pressure resulting in failure of ECC supply and eventually causing core damage. The reliability issues about the performance of passive safety systems have been discussed owing to their high uncertainties, low performance, and lack of experience in operation compared to active safety systems. In aspect of probabilistic safety, complex circuits which comprise many valves and pipelines have possibilities of single failure and common cause failure. Development of innovative passive safety system having differentiated working principle, significant performance, and low possibility of failure can enhance reactor safety providing solutions for the aforementioned problems. Based on these requirements, hybrid control rod which combines the functions of control rod and heat pipe was proposed for the development of passive in-core cooling system (PINCs). The control rods drop to the core using gravity and shutdown the reactor by neutron absorption. The thermosyphon heat pipe is a passive heat transfer device using phase change and convection of working fluid in a closed metal container having two different temperature interfaces (evaporator and condenser). The combination of thermosyphon and control rod, hybrid control rod can achieve reactor shutdown and decay heat removal simultaneously at accident conditions. Hybrid control rod was designed considering the aspects of neutronics (reactivity worth) and mechanical integrity. Most of the nuclear reactors operate at high temperature and high pressure environment with high power density. Thus, pressure control strategies of the hybrid control rod using non-condensable gas and expansion of the working fluid were established to achieve high decay heat removal capacity and operating conditions. The designed hybrid control rods were equipped on the experimental facility and their thermal performances were studied under various amount of working fluid, amount of non-condensable gas, and operating pressures of the test section. The experimental results showed relations between heat transfer characteristics and controlled parameters. Controlling operating condition of hybrid control rod in high pressure worked successfully, and the proportionality between maximum heat removal capacity and operating pressure of hybrid control rod design has been proven. Measured maximum heat transfer rate of single hybrid control rod was 6 kW at 20 bar. Simulations of multi-dimensional analysis for reactor safety (MARS) code were also performed to validate the experimental results and evaluate the prediction capability of the code on the hybrid control rod. The simulation results showed the limits of heat transfer models in the code analyzing the hybrid control rod in which the boiling and condensation heat transfer occurs simultaneously in a manner of countercurrent flow. The experimental results were compared with several models associated with boiling heat transfer, condensation heat transfer, and critical heat flux (CHF) of thermosyphon for the development or the selection of optimal models. The selected models could be implemented to system analysis codes in the purpose of deterministic safety assessment of PINCs against design basis accidents. Imura???s correlation, which was developed in two-phase natural convection condition and validated with experiments in wide range, was selected as boiling heat transfer model of pressurized hybrid control rod. The existing condensation models were based on Nusselt???s film condensation theory. Hence, the effect of non-condensable gas and perturbation between upward vapor flow and downward liquid film flow were not considered at the same time. The change of effective heat transfer length due to presence of non-condensable gas and effect of fluid inertia were considered for the derivation of new condensation model. The main thermal-hydraulic phenomenon which induces CHF of thermosyphon is flooding. The flooding-based CHF models for thermosyphon were derived with theories on instability of the liquid film or maximum liquid film flow rate in countercurrent flow condition. The limited prediction capabilities of the models were attributed to difference between hydraulic diameter and heated diameter as well as high operating pressure. Consequently, new model regarding the CHF of hybrid control rod was suggested to explain its unique characteristics. The hybrid control rod could be equipped on spent fuel dry storage casks for the extension of their thermal margins. The mock-up was designed to be scaled-down to 1/10 of metal dry storage cask developed by NAC. The effect of hybrid control rod on thermal margins of the cask was experimentally studied. The equipment of hybrid control rod with installation of heat sink lid reduced the temperature distributions inside the cask at equal power density condition. Application of hybrid control rod could extend the thermal margin up to 30 %. Feasibility of PINCs based on experimentally and analytically studied hybrid control rods were discussed according to commercial reactors. A number of nuclear facilities has been built to supply and manage energy. The nuclear fuels generate decay heat even in shutdown condition by fission products. Management of the decay heat is important to satisfy demand for nuclear safety. Therefore, new conceptual safety system is required to supplement the issues on existing safety systems. Passive in-core cooling system based on hybrid control rod is the effective way to be applied on extensive nuclear facilities containing nuclear fuels. Pressurized hybrid control rod could meet the operating conditions of application objects with significant decay heat removal capacity.ope

Experimental validation of a mathematical model for the evolution of the particle morphology of waterborne polymer-polymer hybrids: paving the way to the design and implementation of optimal polymerization strategies

Polymer-polymer composite nanoparticles allow both the improvement of the performance in stablished applications of waterborne polymer dispersions and targeting new applications that are out of reach of currently available products. The performance of these materials is determined by the particle morphology. To open the way to process optimization and on-line control of the particle morphology, the capability of the recently developed model to predict the evolution of the particle morphology during seeded semibatch emulsion polymerization process was evaluated. Structured polymer particles were synthesized by copolymerization of styrene and butyl acrylate (St-BA) on methyl methacrylate and butyl acrylate (MMA–BA) copolymer seeds of different Tgs. The model captured well the effect of process variables on the evolution of the particle morphology, opening the way to the design and implementation of optimal strategies.The financial support of the RECOBA project (funding from European Framework Horizon 2020, No. 636820) is gratefully acknowledged