Effect of Noncondensable Gases on Circulation of Primary Coolant in Nuclear Power Plants in Abnormal Situations

The present study focuses on two effects of the presence of a noncondensable gas on the thermal-hydraulic behavior of thecoolant of the primary circuit of a nuclear reactor in the VVER-440 geometry inabnormal situations. First, steam condensation with the presence of air was studied in the horizontal tubes of the steam generator (SG) of the PACTEL test facility. The French thermal-hydraulic CATHARE code was used to study the heat transfer between the primary and secondary side in conditions derived from preliminary experiments performed by VTT using PACTEL. In natural circulation and single-phase vapor conditions, the injection of a volume of air, equivalent to the totalvolume of the primary side of the SG at the entrance of the hot collector, did not stop the heat transfer from the primary to the secondary side. The calculated results indicate that air is located in the second half-length (from the mid-length of the tubes to the cold collector) in all the tubes of the steam generator The hot collector remained full of steam during the transient. Secondly, the potential release of the nitrogen gas dissolved in the water of the accumulators of the emergency core coolant system of the Loviisa nuclear power plant (NPP) was investigated. The author implemented a model of the dissolution and release ofnitrogen gas in the CATHARE code; the model created by the CATHARE developers. In collaboration with VTT, an analytical experiment was performed with some components of PACTEL to determine, in particular, the value of the release time constant of the nitrogen gas in the depressurization conditions representative of the small and intermediate break transients postulated for the Loviisa NPP. Such transients, with simplified operating procedures, were calculated using the modified CATHARE code for various values of the release time constant used in the dissolution and release model. For the small breaks, nitrogen gas is trapped in thecollectors of the SGs in rather large proportions. There, the levels oscillate until the actuation of the low-pressure injection pumps (LPIS) that refill the primary circuit. In the case of the intermediate breaks, most of the nitrogen gas is expelled at the break and almost no nitrogen gas is trapped in the SGs. In comparison with the cases calculated without taking into account the release of nitrogen gas, the start of the LPIS is delayed by between 1 and 1.75 h. Applicability of the obtained results to the real safety conditions must take into accountthe real operating procedures used in the nuclear power plant.

Coolant Flow and Heat Transfer in Pebble-Bed Reactor Core

This thesis gathers knowledge about ongoing high-temperature reactor projects around the world. Methods for calculating coolant flow and heat transfer inside a pebble-bed reactor core are also developed. The thesis begins with the introduction of high-temperature reactors including the current state of the technology. Process heat applications that could use the heat from a high-temperature reactor are also introduced. A suitable reactor design with data available in literature is selected for the calculation part of the thesis. Commercial computational fluid dynamics software Fluent is used for the calculations. The pebble-bed is approximated as a packed-bed, which causes sink terms to the momentum equations of the gas flowing through it. A position dependent value is used for the packing fraction. Two different models are used to calculate heat transfer. First a local thermal equilibrium is assumed between the gas and solid phases and a single energy equation is used. In the second approach, separate energy equations are used for the phases. Information about steady state flow behavior, pressure loss, and temperature distribution in the core is obtained as results of the calculations. The effect of inlet mass flow rate to pressure loss is also investigated. Data found in literature and the results correspond each other quite well, considered the amount of simplifications in the calculations. The models developed in this thesis can be used to solve coolant flow and heat transfer in a pebble-bed reactor, although additional development and model validation is needed for better accuracy and reliability.

Evaluation of the performance of α-SiAlON tool when turning Ti-6Al-4V alloy without coolant

Due to their high hardness and wear resistance, Si3N4 based ceramics are one of the most suitable cutting tool materials for machining cast iron, nickel alloys and hardened steels. However, their high degree of brittleness usually leads to inconsistent results and sudden catastrophic failures. This necessitates a process optimization when machining superalloys with Si3N4 based ceramic cutting tools. The tools are expected to withstand the heat and pressure developed when machining at higher cutting conditions because of their high hardness and melting point. This paper evaluates the performance of α-SiAlON tool in turning Ti-6Al-4V alloy at high cutting conditions, up to 250 m min-1, without coolant. Tool wear, failure modes and temperature were monitored to access the performance of the cutting tool. Test results showed that the performance of α-SiAl0N tool, in terms of tool life, at the cutting conditions investigated is relatively poor due probably to rapid notching and excessive chipping of the cutting edge. These facts are associated with adhesion and diffusion wear rate that tends to weaken the bond strength of the cutting tool.

Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed

Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper head of the reactor vessel as a result of circumferential cracking of a control rod drive mechanism penetration nozzle - which has cast even greater importance on the study of SBLOCAs. Several experimental tests have been performed at the Large Scale Test Facility to simulate the behavior of a PWR during an upper-head SBLOCA. The last of these tests, Organisation for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1, was performed in 2005. This test was simulated with the TRACE 5.0 code, and good agreement with the experimental results was obtained. Additionally, a broad analysis of an upper-head SBLOCA with high-pressure safety injection failed in a Westinghouse PWR was performed taking into account different accident management actions and conditions in order to check their suitability. This issue has been analyzed also in the framework of the OECD/NEA ROSA project and the Code Applications and Maintenance Program (CAMP). The main conclusion is that the current emergency operating procedures for Westinghouse reactor design are adequate for these kinds of sequences, and they do not need to be modified.

Design summary report of LCRE primary coolant pumps and sump /

"Contract AT(30-1)-2789."

Design summary report of LCRE reflector coolant pumps and sump /

"Contract AT(30-1)-2789."

Design summary report of LCRE secondary coolant pump and sump /

"Contract AT(30-1)-2789."

Experiments for CFD-modelling of cooling water and Insulation debris two-phase flow phenomena during loss of coolant accidents

The knowledge of insulation debris generation and transport gains in importance regarding reactor safety research for PWR and BWR. The insulation debris released near the break consists of a mixture of very different fibres and particles concerning size, shape, consistence and other properties. Some fraction of the released insulation debris will be transported into the reactor sump where it may affect emergency core cooling. Experiments are performed to blast original samples of mineral wool insulation material by steam under original thermal-hydraulic break conditions of BWR. The gained fragments are used as initial specimen for further experiments at acrylic glass test facilities. The quasi ID-sinking behaviour of the insulation fragments are investigated in a water column by optical high speed video techniques and methods of image processing. Drag properties are derived from the measured sinking velocities of the fibres and observed geometric parameters for an adequate CFD modelling. In the test rig "Ring line-II" the influence of the insulation material on the head loss is investigated for debris loaded strainers. Correlations from the filter bed theory are adapted with experimental results and are used to model the flow resistance depending on particle load, filter bed porosity and parameters of the coolant flow. This concept also enables the simulation of a particular blocked strainer with CFDcodes. During the ongoing work further results of separate effect and integral experiments and the application and validation of the CFD-models for integral test facilities and original containment sump conditions are expected.

Fukushima-derived radionuclides in sediments of the Japanese Pacific Ocean coast and various Japanese water samples (seawater, tap water, and coolant water of Fukushima Daiichi reactor unit 5)

We investigated Ocean sediments and seawater from inside the Fukushima exclusion zone and found radiocesium (134Cs and 137Cs) up to 800 Bq kg-1 as well as 90Sr up to 5.6 Bq kg-1. This is one of the first reports on radiostrontium in sea sediments from the Fukushima exclusion zone. Seawater exhibited contamination levels up to 5.3 Bq kg-1 radiocesium. Tap water from Tokyo from weeks after the accident exhibited detectable but harmless activities of radiocesium (well below the regulatory limit). Analysis of the Unit 5 reactor coolant (finding only 3H and even low 129I) leads to the conclusion that the purification techniques for reactor coolant employed at Fukushima Daiichi are very effective.

High speed versus conventional grinding in high removal rate machining of alumina and alumina-titania

High removal rate (up to 16.6 mm(3)/s per mm) grinding of alumina and alumina-titania was investigated with respect to material removal and basic grinding parameters using a resin-bond 160 mu m grit diamond wheel at the speeds of 40 and 160 m/s, respectively. The results show that the material removal for the single-phase polycrystalline alumina and the two-phase alumina-titania composite revealed identical mechanisms of microfracture and grain dislodgement under the grinding conditioned selected. There were no distinct differences in surface roughness and morphology for both materials ground at either conventional or high speed. An increase in material removal rate did not necessarily worsen the surface toughness for the two materials at both speeds. Also the grinding forces for the two ceramics demonstrated similar characteristics at any grinding speeds and specific removal rates. Both normal and tangential grinding forces and their force ratios at the high speed were lower than those at the conventional speed, regardless of removal rates. An increase in specific removal rate caused more rapid increases in normal and tangential forces obtained at the conventional grinding speed than those at the high speed. Furthermore, it is found that the high speed grinding at all the removal rates exerted a great amount of coolant-induced normal forces in grinding zone, which were 4-6 times higher than the pure normal grinding forces. (c) 2004 Elsevier Ltd. All rights reserved.

Adhesion after erbium, chromium:yttrium-scandium-gallium-garnet laser application at three different irradiation conditions

The aim of this study was to investigate whether distinct cooling of low fluence erbium, chromium:yttrium-scandium-gallium-garnet (Er,Cr:YSGG) laser irradiation would influence adhesion. Main factors tested were: substrates (two), irradiation conditions (three), and adhesives (three). A 750 mu m diameter tip was used, for 50 s, 1 mm from the surface, with a 0.25 W power output, 20 Hz, energy density of 2.8 J/cm(2) with energy per pulse of 12.5 mJ. When applied, water delivery rate was 11 ml/min. The analysis of variance (ANOVA) showed that laser conditioning significantly decreased the bond strength of all adhesive systems applied on enamel. On dentin, laser conditioning significantly reduced bond strength of etch-and-rinse and one-step self-etch systems; however, laser irradiation under water cooling did not alter bonding of two-step self-etching. It may be concluded that the irradiation with Er,Cr:YSGG laser at 2.8 J/cm(2) with water coolant was responsible for a better adhesion to dentin, while enamel irradiation reduced bond strength, irrespective of cooling conditions.