Frank-Peter Weiss

Influence of the pipe diameter on the structure of the gas-liquid interface in a vertical two-phase pipe flow

Abstract Air-water two-phase flow tests in a large vertical pipe of 194.1-mm inner diameter (i.d.) are reported. Close to the outlet of a 9-m-tall test section, two wire-mesh sensors are installed that deliver instantaneous void fraction distributions over the entire cross section with a resolution of 3 mm and 2500 Hz used for fast-flow visualization. Void fraction profiles, gas velocity profiles, and bubble-size distributions were obtained. A comparison to a small pipe of 52.3-mm i.d. (DN50) revealed significant scaling effects. Here, the increase of the airflow rate leads to a transition from bubbly via slug to churn-turbulent flow. This is accompanied by an appearance of a second peak in the bubble-size distribution. A similar behavior was found in the large pipe; though the large bubbles have a significantly larger diameter at identical superficial velocities, the peak is less high but wider. These bubbles move more freely in the large pipe and show more deformations. The shapes of such large bubbles were characterized in three dimensions. They can be rather complicated and far from ideal Taylor bubbles. Also, the small bubble fraction tends to bigger sizes in the large pipe.

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

Abstract The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold emergency core cooling (ECC) water entering the RPV through the ECC injection into the cold leg. The experimental results show an incomplete mixing with typical concentration and temperature distributions at the core inlet, which strongly depend on the boundary conditions. Computational fluid dynamics calculations were found to be in good agreement with the experiments.

Dynamics of Molten Salt Reactors

Abstract The dynamics of the molten salt reactor (MSR), one of the Generation IV International Forum concepts, was studied. The graphite-moderated channel-type MSR was selected for numerical simulation. MSR, a liquid-fueled reactor, has specific dynamics with two physical peculiarities: The delayed neutron precursors are drifted by the fuel flow, and the fission energy is released directly into the coolant. Presently, there are few accessible numerical codes appropriate for MSR simulation; therefore, the DYN1D-MSR and DYN3D-MSR codes were developed based on the light water reactor dynamics code DYN3D. These allow calculation of one-dimensional and full three-dimensional transient neutronics in combination with parallel channel-type thermal hydraulics. The codes were validated with experimental results of the Molten Salt Reactor Experiment from Oak Ridge National Laboratory and applied to several transients typical for a liquid fuel system. Those transients were initiated by reactivity insertion, by cold or overfueled slugs, by the fuel pump start-up or shutdown, or by the blockage of selected fuel channels. In these considered transients, the response of MSR is characterized by the immediate change of the fuel temperature relative to the temperature at that power level. This causes fast insertion of feedback reactivity, which is negative for power-related temperature increase. On the other hand, the graphite response is slower, and its feedback coefficient depends on the core size and geometry. The addition of erbium to the graphite can ensure negative feedback and inherent safety features also for big low leakage cores. The DYN1D-MSR and DYN3D-MSR codes have been shown to be effective tools for MSR dynamics studies. The MSR response to the majority of transients is considered acceptable within safety margins as long as the graphite feedback coefficient is negative. A transient that is possibly an exception is a local channel blockage.

Experimental and Numerical Investigation of Boron Dilution Transients in Pressurized Water Reactors

Abstract Within the pressurized water reactor (PWR) safety analyses, attention has increasingly focused in recent years on boron dilution events that could potentially lead to reactivity transients. Mixing of the low-boron water with the ambient coolant of higher boron content provides an important mitigation mechanism before the low-boron water enters the core. Experimental support is needed to validate the computational tools to be applied to analyze the mixing of the low-boron water. Experiments were performed in the three test facilities—the Upper Plenum Test Facility (UPTF), the Primärkreislauf (PKL), and the Rossendorf coolant mixing model (ROCOM)—in Germany. The relevant PKL and UPTF tests were focused on small-break loss-of-coolant accident (SBLOCA) scenarios with reflux-condenser mode and restart of natural circulation. The two test facilities represent a typical western-type PWR and are/were operated by Siemens/KWU now Framatome ANP in Germany. While the restart of natural circulation was investigated in the PKL system test facility (volume 1:145, height 1:1), the UPTF experiments dealt with the mixing of water flows with different boron concentration in the cold legs, reactor pressure vessel (RPV) downcomer, and the lower plenum (all these components were full-scale models). The results from the PKL test facility demonstrate that in case of a postulated SBLOCA with reflux condensation phase, natural circulation does not start up simultaneously in all loops. This means that slugs of condensate, which might have accumulated in the pump seal during reflux-condenser mode of operation, would reach the RPV at different points in time. The UPTF tests showed an almost ideal mixing of water flows with different boron concentration in the RPV downcomer. The ROCOM test facility has been built in a linear scale of 1:5 for the investigation of coolant mixing phenomena in a wide range of flow conditions in the RPV of the German KONVOI-type PWR. The test results presented are focused on the mixing of a slug of deborated water during the startup of the first reactor coolant pump. Based on experimentally determined pulse responses, a semianalytical model for the description of coolant mixing inside the KONVOI RPV has been developed. Calculations for a presumed boron dilution event during the startup of the first reactor coolant pump have been carried out by means of the semianalytical model and independently by means of the computational fluid dynamics code CFX-4. The semianalytical model is able to describe the time dependent behavior of the deboration front at each fuel element position in a good agreement with the experiment. All main mixing effects, observed in the experiment, are also reproduced by the CFX calculation.

Use of Zirconium-Based Moderators to Enhance Feedback Coefficients in a MOX-Fueled Sodium-Cooled Fast Reactor

Abstract This work shows the effect of the use of moderating layers on the sodium void effect in sodium-cooled, mixed oxide-fueled fast breeder reactors. The moderating layers consist of either zirconium boride ZrB2 or zirconium hydride ZrH2. The two investigated ZrH2 layers (0.1 and 0.2 mm thick) cause a strong reduction of the sodium void effect. Additionally, these layers significantly improve the fuel temperature effect and the coolant effect of the system. All changes caused by the insertion of the ZrH2 layers result in a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides. The effect in the infinite system can be fully combined with the traditional methods of increasing the neutron leakage.

Influence of System and Neutron-Kinetic Parameter Variations on an Anticipated Transient Without SCRAM in a PWR

The complete failure of the reactor scram system upon request during an operational transient is called anticipated transient without scram (ATWS). According to the German regulatory guidelines, postulated ATWS events have to be analyzed with regard to their consequences on the safety of nuclear power plants. Since the course of ATWS transients is determined by a strong interaction of the neutron kinetics with the thermal hydraulics of the system, coupled 3D neutron kinetic/thermal hydraulic code systems are adequate tools for the analysis of such transients. The coupled code system DYN3D/ATHLET developed at FZD is applied to the analysis of an ATWS transient. The objective of the present work is to perform a best-estimate analysis with consequent use of a 3D neutron kinetic code (DYN3D) in combination with an advanced thermal hydraulic system code (ATHLET) together with a quantification of differences in the course and the results of transients, which arise from the uncertainties of thermal hydraulic and neutron-physical conditions. Typically, the complete failure of the main feed water supply is assumed to be the bounding ATWS event with regard to the maximum primary coolant pressure, which can be reached during the transient. The limitation of the coolant pressure is a precondition for the integrity of the primary circuit. The situation is aggravated if the main coolant pumps remain in operation. For this particular transient, the influence of different thermal hydraulic and neutron-physical conditions on the course of the transient was analyzed. In a number of code runs all systems which have an influence on the course of the transient were varied. These are the auxiliary boration system and the auxiliary feed water supply. Further, the influence of the modeling of the pressurizer safety and relief valves as well as the steam bypass system on the secondary side was assessed. The variation of the pressurizer relief and safety valve behavior has the biggest influence on the primary circuit coolant pressure. In the second part, two different core loading patterns were generated for the analyses by varying the number of MOX (mixed oxide) fuel assemblies (FA) in the core. The basic core loading contains 64 MOX FA. All these MOX FA were replaced by standard uranium oxide FA. The presence of MOX in the core has a remarkable influence on the reactivity coefficients of the fuel temperature and the moderator density. These two parameters mainly influence the behavior of the coolant pressure in the first part of the transient. It has been demonstrated that the pressure maximum decreases with growing MOX portion in the core. The maximum pressure determined in the calculations with variation of system and neutron-physical boundary conditions does not reach the allowed limit for the primary circuit.Copyright

CFD-modeling and experiments of insulation debris transport phenomena in water flow

Abstract The investigation of insulation debris generation, transport, and sedimentation becomes more important with regard to reactor safety research for pressurized water reactors and boiling water reactors when considering the long-term behavior of emergency core coolant systems during all types of loss-of-coolant accidents (LOCAs). The insulation debris released near the break during a LOCA incident consists of a mixture of disparate particle populations that varies with size, shape, consistency, and other properties. Some fractions of the released insulation debris can be transported into the reactor sump, where it may perturb/impinge on the emergency core cooling systems. Open questions of generic interest are, for example, the particle load on strainers and corresponding pressure drop, the sedimentation of the insulation debris in a water pool, and its possible resuspension and transport in the sump water flow. A joint research project on such questions is being performed in cooperation with the University of Applied Sciences Zittau/Görlitz. The project deals with the experimental investigation and the development of computational fluid dynamics (CFD) models for the description of particle transport phenomena in coolant flow. While the experiments are performed at the University of Applied Sciences Zittau/Görlitz, the theoretical work is concentrated at Forschungszentrum Dresden-Rossendorf. In the current paper the basic concepts for CFD modeling are described and feasibility studies including the conceptual design of the experiments are presented.

Verification and Validation of Numerical Models of the Transport of Insulation Debris

A combination of the two-fluid and drift flux models have been used to model the transport of fibrous debris. This debris is generated during loss of coolant accidents in the primary circuit of pressurized or boiling water nuclear reactors, as high pressure steam or water jets can damage adjacent insulation materials including mineral wool blankets. Fibre agglomerates released from the mineral wools may reach the containment sump strainers, where they can accumulate and compromise the long-term operation of the emergency core cooling system. Single-effect experiments of sedimentation in a quiescent rectangular column and sedimentation in a horizontal flow are used to verify and validate this particular application of the multiphase numerical models. The utilization of both modeling approaches allows a number of pseudocontinuous dispersed phases of spherical wetted agglomerates to be modeled simultaneously. Key effects on the transport of the fibre agglomerates are particle size, density and turbulent dispersion, as well as the relative viscosity of the fluid-fibre mixture.

Numerical Simulation of the Insulation Material Transport to a PWR Core Under Loss of Coolant Accident Conditions

In 1992, strainers on the suction side of the ECCS pumps in Barseback NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally-insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modelled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the insulation material during reverse flow. This will certainly further improve the coolability of the core. The spacer grids were modelled as a strainer, which completely retains all the insulation material reaching the uppermost spacer level. There, the accumulation of the insulation material gives rise to the formation of a compressible fibrous cake, the permeability of which to the coolant flow is calculated in terms of the local amount of deposited material and the local value of the superficial liquid velocity. Before the switch over of the ECC injection from the flooding mode to the sump mode, the coolant circulates in an inner convection loop in the core extending from the lower plenum to the upper plenum. The CFD simulations have shown that after starting the sump mode, the ECC water injected through the hot legs flows down into the core at so-called “breakthrough channels” located at the outer core region where the downward leg of the convection roll had established. The hotter, lighter coolant rises in the centre of the core. As a consequence, the insulation material is preferably deposited at the uppermost spacer grids positioned in the breakthrough zones. This means that the fibers are not uniformly deposited over the core cross section. When the inner recirculation stops later in the transient, insulation material can also be collected in other regions of the core. Nevertheless, with a total of 2.7 kg fiber material deposited at the uppermost spacer level, the pressure drop over the fiber cake is not higher than 8 kPa and all the ECC water could still enter the core.Copyright

Hydrodynamic modeling of mineral wool fiber suspensions in a two-dimensional flow

A consequence of a loss of coolant accident is that the local insulation material is damaged and maybe transported to the containment sump where it can penetrate and/or block the sump strainers. An experimental and theoretical study, which examines the transport of mineral wool fibers via single and multi-effect experiments is being performed. This paper focuses on the experiments and simulations performed for validation of numerical models of sedimentation and resuspension of mineral wool fiber agglomerates in a racetrack type channel. Three velocity conditions are used to test the response of two dispersed phase fiber agglomerates to two drag correlations and to two turbulent dispersion coefficients. The Eulerian multiphase flow model is applied with either one or two dispersed phases.