Thomas Schulenberg

Design analysis of core assemblies for supercritical pressure conditions

The increase of steam parameters to supercritical conditions could reduce the power generating costs of light water reactors significantly [Proceedings of SCR-2000 (2000) 1]. Core assemblies, however, will differ from current BWR or PWR design. In this context, this paper summarizes the main results related to a thermal-hydraulic design analysis of applicable fuel assemblies. Starting from a thorough literature survey on heat transfer of supercritical fluids, the current status indicates a large deficiency in the prediction of the heat transfer coefficient under reactor prototypical conditions. For the thermal-hydraulic design of such fuel assemblies the sub-channel analysis code Sub-channel Thermal-hydraulic Analysis in Fuel Assemblies under Supercritical conditions (STAFAS) has been developed, which will have a higher numerical efficiency compared to the conventional sub-channel analysis codes. The effect of several design parameters on the thermal-hydraulic behaviour in sub-channels has been investigated. Based on the results achieved so far, two fuel assembly configurations are recommended for further design analysis, i.e. a tight square lattice and a semi-tight hexagonal lattice.

High Performance Light Water Reactor : Design and Analyses

Sc h u le n b er g // S ta rfl in g er (e d s. ) HIGH PERFORMANCE LIGHT WATER REACTOR Design and Analyses The High Performance Light Water Reactor is a nuclear reactor concept of the 4th generation which is cooled and moderated with supercritical water. The concept has been worked out by a consortium of European industry, research centers and universities, co-funded by the European Commission. It features a once through steam cycle, a pressure vessel type reactor, and a compact containment with pressure suppression pool. The conceptual design described here shall enable to assess its feasibility, its safety features and its economic potential. Thomas Schulenberg // Jörg Starflinger (eds.)

Flow Instability and Critical Heat Flux for Flow Boiling of Water in a Vertical Annulus at Low Pressure

We investigated the critical heat flux (CHF) for flow boiling of water in a vertical annulus. The coaxial annulus has a diameter ratio of 1.37 and the inner zircaloy tube is heated directly over a length of 325 mm. CHF can occur prematurely due to flow instabilities. Therefore, we analyzed the flow stability at different heat input conditions using two types of pumps, a rotary and a gear type pump. The unstable CHF occurred at 61% and 90% of the stable value for the rotary and the gear type pump, respectively. Consequently, the following CHF experiments were conducted at stable flow conditions. The outlet pressure was constant at 120 kPa, the mass flux varied from 250 to 1000 kg/(m2 s) and the inlet subcooling was at 102, 167, and 250 kJ/kg. The CHF results increase with mass flux from 0.67 to 2.62 MW/m2 and show similar trends compared to literature data. However, the experimental data for flow boiling in annuli at low pressure are limited. Additionally, we measured the dynamic contact angle between the zircaloy tube surface and water using the Wilhelmy method.Copyright

Turbulence Structures in Horizontal Two-Phase Flows Under Counter-Current Conditions

In order to improve the multi-dimensional numerical simulation of horizontal two-phase flows, the knowledge of local turbulent quantities is of great importance. In horizontal stratified flows, the denser (first) phase flows as a film beneath the other (second) phase. Under counter-current conditions, the second phase flows into the opposite direction of the first phase. In the present investigations a liquid film flows counter-currently to a gas flow. According to the flow rates of both phases, different flow regimes set in. In supercritical flows (Fr>1), the height of the liquid film increases in flow direction, while it decreases in subcritical flows (Fr<1). For sufficiently high gas flow rates the upper part of the liquid film flows into direction of the gas flow, while the lower part still flows into its initial direction opposite to the gas flow. Only a reduced amount of water reaches the end of the test section. This flow regime is referred to as partially reversed flow. The presented local measurements provide not only the mean and rms-velocities of the liquid film, but also the corresponding Reynolds stresses. Local measurements are carried out at two different positions along the test section for various boundary conditions. Furthermore, the liquid injection height has been varied. The kinematic and turbulent structures of the different flow patterns are presented and compared.Copyright

Transient Heat Transfer in an Out-of-Pile SCWR Fuel Assembly Test at Near-Critical Pressure

While supercritical water is a perfect coolant with excellent heat transfer, a temporary decrease of the system pressure to subcritical conditions, either during intended transients or by accident, can easily cause a boiling crisis with significantly higher cladding temperatures of the fuel assemblies. These conditions have been tested in an out-of-pile experiment with a bundle of four heated rods in the supercritical water multipurpose loop (SWAMUP) facility coconstructed by CGNPC and SJTU in China. Some of the transient tests have been simulated at KIT with a one-dimensional (1D) MATLAB code, assuming quasi-steady-state flow conditions, but time dependent temperatures in the fuel rods. Heat transfer at supercritical and at near-critical conditions was modeled with a recent look-up table of Zahlan (2015, “Derivation of a Look-Up Table for Trans-Critical Heat Transfer in Water Cooled Tubes,” Ph.D. dissertation, University of Ottawa, Ottawa, ON, Canada.), and subcritical film boiling was modeled with the look-up table of Groeneveld et al. (2003, “A Look-Up Table for Fully Developed Film Boiling Heat Transfer,” Nucl. Eng. Des., 225(1), pp. 83–97.). Moreover, a conduction controlled rewetting process was included in the analyses, which is based on an analytical solution of Schulenberg and Raqu e (2014, “Transient Heat Transfer During Depressurization From Supercritical Pressure,” Int. J. Heat Mass Transfer, 79(12), pp. 233–240.). The method could well reproduce the boiling crisis during depressurization from supercritical to subcritical pressure, including rewetting of the hot zone within some minutes, but the peak temperature was somewhat under-predicted. Tests with a lower heat flux, which did not cause such phenomena, could be predicted as well. In another test with increasing pressure, however, a boiling crisis was also observed at a heat flux, which was significantly lower than the critical heat flux (CHF) predicted by the CHF look-up table of Groeneveld et al. (2007, “The 2006 CHF Look-Up Table,” Nucl. Eng. Des., 237(15–17), pp. 1909–1922.). The paper is summarizing the physical models and the numerical approach. Comparison with experimental data is used to discuss the applicability of the method for the design of supercritical water-cooled reactors (SCWR). [DOI: 10.1115/1.4038061]

Expected Safety Performance of the Supercritical Water Reactor Fuel Qualification Test

The supercritical water reactor (SCWR) fuel qualification test is an in-pile test of a four-rod fuel assembly at supercritical pressure inside a research reactor, which is operated at atmospheric pressure. The risk of radioactive release from this new test facility should not exceed the accepted risk of the existing research reactor. A large number of safety analyses have been performed to assess this risk, which are summarized in this paper. Among them are studies of design basis accidents, assuming different failure modes of the high-pressure system, as well as an assessment of consequences of postulated accidents beyond the design basis. Results show that the safety objectives can be met.

Design of an In-Pile SCWR Fuel Qualification Test Loop

In the development of the supercritical water cooled reactor (SCWR), an in-pile fuel assembly test loop has been designed within the framework of the joint Chinese-European project, called SCWR-FQT (Fuel Qualification Test). This paper presents basic design of the loop with its auxiliary and safety systems, which has been examined in detail by thermal hydraulic analyses in order to achieve operation of the loop above the thermodynamic critical point of water (374 °C, 22.1 MPa) and checked by stress analyses to assure safe operation.The designed experimental loop for fuel qualification in supercritical water consists of a closed pressurized water circuit with forced circulation of the coolant through the test section - the active channel which is intended to be installed into the existing research pool type reactor LVR-15. The active channel will be operated at temperatures and pressures, which are typical for the High Performance Light Water Reactor (HPLWR). A thick-walled pressure tube made from austenitic stainless steel able to withstand the high system pressure, encloses the active channel. It contains four fuel rods with UO2 (enrichment of 19.7% U235) with a total heating power of ~64 kW and a recuperator in order to achieve hot channel conditions as they are expected to occur in the evaporator of the HPLWR. The internal flow is realized so as to prevent the creep condition of the pressure tube. An internal U-tube cooler serves as heat sink and is connected to the secondary circuit. The entire active channel is isolated from water of the reactor pool by an air gap between the pressure tube and an aluminum displacer. The test section with fuel is connected to a 300°C closed loop and to a primary pump located outside the reactor building as well as safety systems and auxiliary systems such as purification and measurement circuits, which are all connected with the primary circuit.keywords: design, SCWR, Generation-IV, thermal hydraulic, SCWR-FQT

A hard optimisation test function with symbolic solution visualisation for fast interpretation by the human eye

We propose a class of test problems for evaluating the performance of global function optimisers based on finding an optimal spatial distribution of nonidentical particles interacting with two different potential fields. Because of the possibility of intuitive solution visualisation it can be of particular benefit during development of optimisation algorithms. An ensemble of N particles is constrained to a low-dimensional space and each particle contributes in two ways to the total potential energy: by its position on a hilly track and through repulsive neighbour potentials. The task of minimising the ensembles total potential energy corresponds to searching an N-dimensional space with many local minima separated through higher and lower barriers; hence, it can serve as a performance measure for evolutionary algorithms (EA). The search difficulty is scalable through the number of particles and the hilliness of the track. In particular, if the particles are made nonidentical by giving them different masses or charges, the search will become very challenging because of the introduced combinatorial aspect and the “curse of dimensionality”. Among many similarly challenging optimisation problems this test function class has the advantage that solution candidates can be plotted in ways which allow humans to estimate not only relative objective function values but also DNA vector relations upon a quick glance. For the EA developer this allows a fast feedback cycle between a modification to the EA and the observed change in optimisation history behaviour. This makes experimentation with EA elements at a fundamental level easier. Furthermore, this class of real-domain search offers a wide range of difficulty and complexity levels and can be split up into a two-objective optimisation.

Validation of the System Code APROS for Fast Transients

The thermal-hydraulic system code APROS Version 5.09 [1] is being applied in the European project SCWR-FQT to evaluate the performance of the safety systems for a nuclear test facility operated with supercritical water. In order to validate the commercial code for predictions of transient phenomena, two adequate hydraulic experiments from literature have been simulated.The experiment of Fujii and Akagawa [2] investigated hydraulic shocks as they will occur in case a pipe, which is stationary passed through by water, is abruptly closed. A simple numerical model with adapted time and space nodalization was able to reproduce the observed physical phenomena, such as the magnitude of the initial pressure wave and reflection time, in detail.In a second experiment, a test series was performed by Becker et al. [3] and Mathisen [4] in order to examine the natural circulation in a closed loop for different heating rates and system pressures. For a stepwise power increase, the typical mass flow characteristic for boiling channels was recorded until the onset of flow oscillations. In further runs, the effect of different initial conditions on the flow stability was analyzed. This paper compares the numerical predictions with both experimental results. The numerical models could describe the physical phenomena with appropriate accuracy.Copyright

Enhancement of heat transfer in fuel assemblies of high performance light water reactors

Heat transfer in fuel assemblies for a High Performance Light Water Reactor can be achieved either with artificial surface roughness of the fuel claddings or by a spiral cross flow between the fuel pins, for which purpose a staircase type grid spacer has been designed. An application of earlier test results with rough claddings for gas cooled reactors to supercritical water conditions, together with new heat transfer estimates for a spiral flow, indicates that the heat transfer coefficient of the coolant at the cladding surface can be increased by more than a factor of two, which will reduce the peak cladding temperature by at least 50°C. This improvement shall allow either to increase the core outlet temperature at a given cladding temperature or to reduce the peak temperature at the envisaged core outlet temperature. The paper includes analyses and design details for realization of such enhancement of heat transfer.