Abstract: In this paper, the results of experimental tests
performed on a Helical Wire Rope Isolator (HWRI) are presented in
order to describe the dynamic and static behavior of the selected
metal device in three different displacements ranges, namely small,
relatively large, and large displacements ranges, without and under
the effect of a vertical load. A testing machine, allowing to apply
horizontal displacement or load histories to the tested bearing with a
constant vertical load, has been adopted to perform the dynamic and
static tests. According to the experimental results, the dynamic
behavior of the tested device depends on the applied displacement
amplitude. Indeed, the HWRI displays a softening and a hardening
stiffness at small and relatively large displacements, respectively, and
a stronger nonlinear stiffening behavior at large displacements.
Furthermore, the experimental tests reveal that the application of a
vertical load allows to have a more flexible device with higher
damping properties and that the applied vertical load affects much
less the dynamic response of the metal device at large displacements.
Finally, a decrease in the static to dynamic effective stiffness ratio
with increasing displacement amplitude has been observed.
Abstract: A Multi-dimensional computational fluid dynamics
(CFD) two-phase model was developed with the aim to simulate
the in-core coolant circuit of a pressurized heavy water reactor
(PHWR) of a commercial nuclear power plant (NPP). Due to the
fact that this PHWR is a Reactor Pressure Vessel type (RPV),
three-dimensional (3D) detailed modelling of the large reservoirs of
the RPV (the upper and lower plenums and the downcomer) were
coupled with an in-house finite volume one-dimensional (1D) code
in order to model the 451 coolant channels housing the nuclear fuel.
Regarding the 1D code, suitable empirical correlations for taking into
account the in-channel distributed (friction losses) and concentrated
(spacer grids, inlet and outlet throttles) pressure losses were used.
A local power distribution at each one of the coolant channels
was also taken into account. The heat transfer between the coolant
and the surrounding moderator was accurately calculated using a
two-dimensional theoretical model. The implementation of subcooled
boiling and condensation models in the 1D code along with the use
of functions for representing the thermal and dynamic properties of
the coolant and moderator (heavy water) allow to have estimations
of the in-core steam generation under nominal flow conditions for a
generic fission power distribution. The in-core mass flow distribution
results for steady state nominal conditions are in agreement with the
expected from design, thus getting a first assessment of the coupled
1/3D model. Results for nominal condition were compared with
those obtained with a previous 1/3D single-phase model getting more
realistic temperature patterns, also allowing visualize low values of
void fraction inside the upper plenum. It must be mentioned that the
current results were obtained by imposing prescribed fission power
functions from literature. Therefore, results are showed with the aim
of point out the potentiality of the developed model.
Abstract: Most HWRs currently use natural uranium fuel. Using enriched uranium fuel results in a significant improvement in fuel cycle costs and uranium utilization. On the other hand, reactivity changes of HWRs over the full range of operating conditions from cold shutdown to full power are small. This reduces the required reactivity worth of control devices and minimizes local flux distribution perturbations, minimizing potential problems due to transient local overheating of fuel. Analyzing heavy water effectiveness on neutronic parameters such as enrichment requirements, peaking factor and reactivity is important and should pay attention as primary concepts of a HWR core designing. Two nuclear nuclear reactors of CANDU-type and hexagonal-type reactor cores of 33 fuel assemblies and 19 assemblies in 1.04 P/D have been respectively simulated using MCNP-4C code. Using heavy water and light water as moderator have been compared for achieving less reactivity insertion and enrichment requirements. Two fuel matrixes of (232Th/235U)O2 and (238/235U)O2 have been compared to achieve more economical and safe design. Heavy water not only decreased enrichment needs, but it concluded in negative reactivity insertions during moderator density variations. Thorium oxide fuel assemblies of 2.3% enrichment loaded into the core of heavy water moderator resulted in 0.751 fission to absorption ratio and peaking factor of 1.7 using. Heavy water not only provides negative reactivity insertion during temperature raises which changes moderator density but concluded in 2 to 10 kg reduction of enrichment requirements, depend on geometry type.
Abstract: An approach is offered for more precise definition of base lines- borders in handwritten cursive text and general problems of handwritten text segmentation have also been analyzed. An offered method tries to solve problems arose in handwritten recognition with specific slant or in other words, where the letters of the words are not on the same vertical line. As an informative features, some recognition systems use ascending and descending parts of the letters, found after the word-s baseline detection. In such recognition systems, problems in baseline detection, impacts the quality of the recognition and decreases the rate of the recognition. Despite other methods, here borders are found by small pieces containing segmentation elements and defined as a set of linear functions. In this method, separate borders for top and bottom border lines are found. At the end of the paper, as a result, azerbaijani cursive handwritten texts written in Latin alphabet by different authors has been analyzed.
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: A total of 33,680 nuclear power plants (NPPs) workers were monitored and recorded from 1990 to 2007. According to the record, the average individual radiation dose has been decreasing continually from it 3.20 mSv/man in 1990 to 1.12 mSv/man at the end of 2007. After the International Commission on Radiological Protection (ICRP) 60 recommendation was generalized in South Korea, no nuclear power plant workers received above 20 mSv radiation, and the numbers of relatively highly exposed workers have been decreasing continuously. The age distribution of radiation workers in nuclear power plants was composed of mainly 20-30- year-olds (83%) for 1990 ~ 1994 and 30-40-year-olds (75%) for 2003 ~ 2007. The difference in individual average dose by age was not significant. Most (77%) of NPP radiation exposures from 1990 to 2007 occurred mostly during the refueling period. With regard to exposure type, the majority of exposures were external exposures, representing 95% of the total exposures, while internal exposures represented only 5%. External effective dose was affected mainly by gamma radiation exposure, with an insignificant amount of neutron exposure. As for internal effective dose, tritium (3H) in the pressurized heavy water reactor (PHWR) was the biggest cause of exposure.