Abstract: Tritium activity concentration in Danube river water
in Serbia has been determinate using a liquid scintillation counter
Quantulus 1220. During December 2010, water samples were taken
along the entire course of Danube through Serbia, from Hungarian-
Serbian to Romanian-Serbian border. This investigation is very
important because of the nearness of nuclear reactor Paks in
Hungary. Sample preparation was performed by standard test method
using Optiphase HiSafe 3 scintillation cocktail. We used a rapid
method for the preparation of environmental samples, without
electrolytic enrichment.
Abstract: The main objective developed in this paper is to find a
graphic technique for modeling, simulation and diagnosis of the
industrial systems. This importance is much apparent when it is about
a complex system such as the nuclear reactor with pressurized water
of several form with various several non-linearity and time scales. In
this case the analytical approach is heavy and does not give a fast
idea on the evolution of the system. The tool Bond Graph enabled us
to transform the analytical model into graphic model and the
software of simulation SYMBOLS 2000 specific to the Bond Graphs
made it possible to validate and have the results given by the
technical specifications. We introduce the analysis of the problem
involved in the faults localization and identification in the complex
industrial processes. We propose a method of fault detection applied
to the diagnosis and to determine the gravity of a detected fault. We
show the possibilities of application of the new diagnosis approaches
to the complex system control. The industrial systems became
increasingly complex with the faults diagnosis procedures in the
physical systems prove to become very complex as soon as the
systems considered are not elementary any more. Indeed, in front of
this complexity, we chose to make recourse to Fault Detection and
Isolation method (FDI) by the analysis of the problem of its control
and to conceive a reliable system of diagnosis making it possible to
apprehend the complex dynamic systems spatially distributed applied
to the standard pressurized water nuclear reactor.
Abstract: The two-phase flow field and the motion of the free
surface in an oscillating channel are simulated numerically to assess
the methodology for simulating nuclear reacotr thermal hydraulics
under seismic conditions. Two numerical methods are compared: one
is to model the oscillating channel directly using the moving grid of
the Arbitrary Lagrangian-Eulerian method, and the other is to simulate
the effect of channel motion using the oscillating acceleration acting
on the fluid in the stationary channel. The two-phase flow field in the
oscillating channel is simulated using the level set method in both
cases. The calculated results using the oscillating acceleration are
found to coinside with those using the moving grid, and the theoretical
back ground and the limitation of oscillating acceleration are discussed.
It is shown that the change in the interfacial area between liquid and
gas phases under seismic conditions is important for nuclear reactor
thermal hydraulics.
Abstract: In Korea, the technology of a load fo nuclear power plant has been being developed.
automatic controller which is able to control temperature and axial power distribution was developed. identification algorithm and a model predictive contact former transforms the nuclear reactor status into
numerically. And the latter uses them and ge
manipulated values such as two kinds of control ro
this automatic controller, the performance of a coperation was evaluated. As a result, the automatic generated model parameters of a nuclear react to nuclear reactor average temperature and axial power the desired targets during a daily load follow.
Abstract: In a nuclear reactor Loss of Coolant accident (LOCA)
considers wide range of postulated damage or rupture of pipe in the
heat transport piping system. In the case of LOCA with/without
failure of emergency core cooling system in a Pressurised Heavy
water Reactor, the Pressure Tube (PT) temperature could rise
significantly due to fuel heat up and gross mismatch of the heat
generation and heat removal in the affected channel. The extent and
nature of deformation is important from reactor safety point of view.
Experimental set-ups have been designed and fabricated to simulate
ballooning (radial deformation) of PT for 220 MWe IPHWRs.
Experiments have been conducted by covering the CT by ceramic
fibers and then by submerging CT in water of voided PTs. In both
the experiments, it is observed that ballooning initiates at a
temperature around 665´┐¢C and complete contact between PT and
Caldaria Tube (CT) occurs at around 700´┐¢C approximately. The
strain rate is found to be 0.116% per second. The structural integrity
of PT is retained (no breach) for all the experiments. The PT heatup
is found to be arrested after the contact between PT and CT, thus
establishing moderator acting as an efficient heat sink for IPHWRs.
Abstract: Thermal behavior of fuel channel under loss of coolant accident (LOCA) is a major concern for nuclear reactor safety. LOCA along with failure of emergency cooling water system (ECC) may leads to mechanical deformations like sagging and ballooning. In order to understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of Indian Pressurized Heavy Water Reactor (IPHWR) under symmetrical and asymmetrical heat-up conditions. For simulating the fully voided scenario, symmetrical heating of pressure was carried out by injecting 13.2 KW (2 % of nominal power) to all the 19 pins and the temperatures of pressure tube, calandria tube and clad tubes were measured. During symmetrical heating the sagging of fuel channel was initiated at 460 °C and the highest temperature attained by PT was 650 °C . The decay heat from clad tubes was dissipated to moderator mainly by radiation and natural convection. The highest temperature of 680 °C was observed over the outer ring of clad tubes of fuel simulator. Again, to simulate partially voided condition, asymmetrical heating of pressure was carried out by supplying 8.0 kW power to upper 8 pins of fuel simulator and temperature profiles were measured. Along the circumference of pressure tube (PT) the highest temperature difference of 320 °C was observed, which highlights the magnitude of thermal stresses under partially voided conditions.
Abstract: The turbulent mixing of coolant streams of different
temperature and density can cause severe temperature fluctuations in
piping systems in nuclear reactors. In certain periodic contraction
cycles these conditions lead to thermal fatigue. The resulting aging
effect prompts investigation in how the mixing of flows over a sharp
temperature/density interface evolves. To study the fundamental
turbulent mixing phenomena in the presence of density gradients,
isokinetic (shear-free) mixing experiments are performed in a square
channel with Reynolds numbers ranging from 2-500 to 60-000.
Sucrose is used to create the density difference. A Wire Mesh Sensor
(WMS) is used to determine the concentration map of the flow in the
cross section. The mean interface width as a function of velocity,
density difference and distance from the mixing point are analyzed
based on traditional methods chosen for the purposes of
atmospheric/oceanic stratification analyses. A definition of the
mixing layer thickness more appropriate to thermal fatigue and based
on mixedness is devised. This definition shows that the thermal
fatigue risk assessed using simple mixing layer growth can be
misleading and why an approach that separates the effects of large
scale (turbulent) and small scale (molecular) mixing is necessary.
Abstract: Stainless steel has been employed in many
engineering applications ranging from pharmaceutical equipment to
piping in the nuclear reactors and storage to chemical products. In
this attempt, simulation of fatigue crack growth based on
experimental results of austenitic stainless steel 304L was presented
using AFGROW code when NASGRO mode laws adopted. Double
through crack at hole specimen is used in this investigation under
constant amplitude loading. Effect of mean stress is highlighted.
Results show that fatigue crack growth rate (FCGR) and fatigue life
were affected by maximum applied load and dimension of hole. An
equivalent of Paris law for this material was estimated.
Abstract: Whereas in the third generation nuclear reactors,
dimensions of core and also the kind of coolant and enrichment
percent of fuel have significantly changed than the second
generation, therefore in this article the aim is based on a
comparative investigation between two same power reactors of
second and third generations, that the neutronic parameters of both
reactors such as: K∞, Keff and its details and thermal hydraulic
parameters such as: power density, specific power, volumetric heat
rate, released power per fuel volume unit, volume and mass of clad
and fuel (consisting fissile and fertile fuels), be calculated and
compared together. By this comparing the efficiency and
modification of third generation nuclear reactors than second
generation which have same power can be distinguished.
In order to calculate the cited parameters, some information
such as: core dimensions, the pitch of lattice, the fuel matter, the
percent of enrichment and the kind of coolant are used. For
calculating the neutronic parameters, a neutronic program entitled:
SIXFAC and also related formulas have been used. Meantime for
calculating the thermal hydraulic and other parameters, analytical
method and related formulas have been applied.
Abstract: We present the development of a new underwater laser
cutting process in which a water-jet has been used along with the
laser beam to remove the molten material through kerf. The
conventional underwater laser cutting usually utilizes a high pressure
gas jet along with laser beam to create a dry condition in the cutting
zone and also to eject out the molten material. This causes a lot of gas
bubbles and turbulence in water, and produces aerosols and waste
gas. This may cause contamination in the surrounding atmosphere
while cutting radioactive components like burnt nuclear fuel. The
water-jet assisted underwater laser cutting process produces much
less turbulence and aerosols in the atmosphere. Some amount of
water vapor bubbles is formed at the laser-metal-water interface;
however, they tend to condense as they rise up through the
surrounding water. We present the design and development of a
water-jet assisted underwater laser cutting head and the parametric
study of the cutting of AISI 304 stainless steel sheets with a 2 kW
CW fiber laser. The cutting performance is similar to that of the gas
assist laser cutting; however, the process efficiency is reduced due to
heat convection by water-jet and laser beam scattering by vapor. This
process may be attractive for underwater cutting of nuclear reactor
components.