Section Reports: 2012 Annual Meeting on Nuclear Technology - Part 5


Section Reports: 2012 Annual Meeting on Nuclear Technology - Part 5

Stieglitz, R.; Mull, T.; Höhne, T.; Rieger, U.; Buettner, K.

The session Testing Plants was chaired by Thomas Höhne (Helmholtz-Zentrum Dresden-Rossendorf e.V.). The first presentation from S. Schollenberger (Co-authors: K. Umminger, B. Schoen, all Areva NP GmbH) was about „PKL-Experiments on failure of RHRS under 3/4-loop operation for closed primary circuit – Phenomena and operational aspects“. At first, S. Schollenberger gave a few introducing remarks on the PKL test facility, he explicated that the PKL (German acronym for Primärkreislauf, primary circuit) is a scaled down replication (1:1 heights, volumes, power input and mass flows of safety and operational systems 1:145) of the nuclear steam supply system of a 1300 MW KWU type pressurized water reactor (PWR). It is operated at AREVA NP Germany to experimentally investigate the thermal hydraulic behavior of PWRs under accident conditions for design and beyond-design basis accidents (e.g. efficiency of accident management measures). He then stated the motivation for the adoption of the topic “failure of RHRS under cold shut-down states in the PKL experiments to be founded in the significant contribution to the integral core damage frequency (CDF) of 1.1∙10-5 p.a. and plant implied by failure of RHRS: broken down to 24 h, a day under shut-down conditions implies a much higher risk than a full load day. He further stressed, that in the long run, a stationary state (stationary boiling) with constant primary pressure and assured heat transport to secondary side (core cooling) becomes always established as long as at least 1 of 4 SGs is operable (MS/ FW available), even without intervention of the operating personnel. He then summarized, that additional coolant injections result in a deterioration of heat transport to secondary and give rise to the RCS pressure level. He concluded that additional coolant injections should therefore only be considered as a preparative measure for the immediate re-connection of the RHRS (operability assumed). S. Schollenbegrer added that heat transfer mechanisms with prospect of a continuous boron dilution in below SG outlet (due to condensate transport across the SGs) may be excluded for the operation of 2 active SGs (out of 4). S. Schollenberger then focused on the operational aspects, he presented recommendations for the reconnection of the RHRS (re-availability assumed): The PKL tests proved the deployment of a single LPSI/RHRS pump in flooding mode (simultaneous hot and cold side injection with maximum flow rate) to be an adequate preparative measure to re-establish operating conditions of the RHRS from boiling condition in the RCS (i.e. quick and enduring subcooling of core, downcomer and hot sides). If only one RHRS line is operational, the flooding (simultaneous hot/cold leg injection) of large amounts of ECC with the available line and the following re-start of RHRS operation after switchover in the same line may allow a safe reconnection of the RHRS.

J. Dumond (AREVA NP GmbH, Offenbach, co-authors: F. Maisberger, AREVA NP GmbH, A. Class, Karlsruhe Institute of Technology, Karlsruhe) presented the “Development and first experimental results of the KERENA Passive Outflow Reducer”. Increased safety and reduced costs are achieved in the boiling water reactor KERENA with a smart combination of active and passive safety systems. One of these passive systems is the Passive Outflow Reducer (POR). It is positioned in the reactor nozzle to passively limit without moving parts the loss of coolant (LOC) following the break of a large pipe connected to the reactor pressure vessel at a low elevation in one flow direction; while it minimizes the flow resistance in the other flow direction. An innovative design consisting of 37 parallel profiled channels where each channel is composed of two Venturi-nozzles was introduced. The development of this design with numerical tools was explained and the experimental validation in the LOC flow direction was described. The experiments were performed for single Venturi-nozzle and double Venturi-nozzle designs under realistic plant conditions. One profiled channel was tested at a time. Experimental results indicated that double-nozzle designs do reduce loss of coolant in comparison to single-nozzle designs (50% for the LOC considered here in the KERENA reactor) and that the innovative POR design meets KERENA requirements. Finally, computational fluid dynamic is developed to further optimize the geometry. In the summary, it was emphasized that this simple innovative passive fluidic diode without moving part is ready for similar applications.
Further Mr. M. Majed (Co-authors: S. Andersson, F. Waldemarsson, all Westinghouse Electric Sweden) showed the presentation „Westinghouse Critical Heat Flux Test Facility – ODEN“. The ODEN test facility is a replacement to (and improvement upon) the well known former Heat Transfer Research Facility (HTRF) of Columbia University in New York City. The ODEN loop shares the laboratory infrastructure (power supply, heat sink and control room) with the well-known FRIGG Boiling Water Reactor (BWR) test loop. The ODEN loop is designed to cover DNB (Departure of Nuclear Boiling) testing needs for all types of PWR lattices in 5x5 or 6x6 rectangular geometry or in hexagonal test sections. The loop installation was completed in 2006, shakedown testing in 2009, and qualification / benchmark testing versus HTRF data was completed in 2010. The ODEN critical heat flux test loop has been utilized recently to perform DNB measurements on Westinghouse fuel design for VVER 1000 type reactors. The test bundle configuration is a 19 rod hexagonal array. The fuel has been tested in an extensive thermal-hydraulic verification program with axially uniform test (typical cell) and two axially cosine tests (typical and thimble cells). The DNB measurements have been performed at low to high pressures (10 to 17 MPa), low to high mass flows, (0.8 to 7.8 kg/s) and include very high mass steam quality conditions, up to and exceeding 50 %. Mr. Majed closed his presentation with the statement that the ODEN loop has again showed the high DNB data quality, and excellent consistency and repeatability of the DNB data were achieved.
The last presentation was given by A. Onea (Co-authors: M. Lux, W. Hering, all Karlsruher Institut für Technologie – KIT) and about „Optimization of the cold trap design for the KASOLA sodium facility“. The authors envisaged the employment of the new sodium facility for research of transmutation, accelerator target development, as well as an innovative assignment of sodium for solar applications. It is reported that the sodium purification will be performed by cooling it below its saturation temperature in a cold trap that integrates in the upper part a Na-air heat exchanger and a Na-Na heat recuperator, while the bottom part integrates a wire mesh, where the main impurities NaH and Na2O can deposit on its “cold” surface. The talk was focused on the optimization and layout of the wire mesh and of the Na-air heat exchanger and Na-Na heat recuperator, for which two design proposals are compared, namely a simple design, in which the sodium exists through a straight vertical pipe that serves as a heat recuperator and an optimized design, in which the sodium exists through an optimized helical coil that serves also as a heat recuperator. The authors discussed the optimal wire mesh capacity for capturing of the impurities, their residence time distribution in the wire mesh, the pressure loss in the wire mesh and the planned operating range of the cold trap. The optimized design offers increased heat transfer surface for both the air cooling circuit and the recuperation side, while the disadvantages of this design are moderate increase in the pressure loss and an increased manufacturing cost. Analyses with ANSYS CFX 13 for the flow dynamics and heat transfer, using a conjugate heat transfer approach and modeling the wire mesh as a porous domain confirmed the expected results: in the optimized design the wire mesh is cooled at least 15°C more than in the simple design, at the same air cooling capacity and furthermore, quasi the entire heat of the sodium is recuperated, denoting a heat recuperation efficiency of ~99% for the entire range of the air flow rate. Due to the enhanced cooling capacity it is reported that the optimized design offers an increased temperature operation range for sodium, while the safe operation is still warranted and although the manufacturing costs are higher as for the simple design, the operating costs will be smaller due to the enhanced cooling.

Keywords: Jahrestagung Kerntechnik; Sektion 2; Technische SItzung

  • atw - International Journal for Nuclear Power 58(2013)1, 43-44
    ISSN: 1431-5254

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