Abstract: A customized two-stage model has been developed to simulate, analyse, and visualize distillation of actinides as a useful alternative low-pressure separation method in the nuclear recycling cases. Under the most optimal conditions of idealized thermodynamic equilibrium stages and under total reflux of distillate the investigated cases of chloride systems for the separation of such actinides are (A) UCl4-CsCl-PuCl3 and (B) ThCl4-NaCl-PuCl3. Simulatively, uranium tetrachloride in case A is successfully separated by distillation into a six-stage distillation column, and thorium tetrachloride from case B into an eight-stage distillation column. For this, a permissible mole fraction value of 1E-06 has been assumed for the residual impurification degree. With further separation effort of eleven to seventeen required separation stages, the monochlorides of plutonium trichloride from both systems A and B are simulatively shown to be separated as high pure distillation products.
Abstract: In this research, we have focused on numeral simulation of a fuel rod in order to examine distribution of heat temperature in components of fuel rod by Fluent software by providing steady state, single phase fluid flow, frequency heat flux in a fuel rod in nuclear reactor to numeral simulation. Results of examining different layers of a fuel rod consist of fuel layer, gap, pod, and fluid cooling flow, also examining thermal properties and fluids such as heat transition rate and pressure drop. The obtained results through analytical method and results of other sources have been compared and have appropriate correspondence. Results show that using heavy water as cooling fluid along with few layers of gas and pod have the ability of reducing the temperature from above 300 ◦C to 70 ◦C. This investigation is developable for any geometry and material used in the nuclear reactor.
Abstract: In the Japan Atomic Energy Agency, three types of experimental research, advanced reactor fuel reprocessing, radioactive waste disposal, and nuclear fuel cycle technology, have been carried out at the Chemical Processing Facility. The facility has generated high level radioactive liquid and solid wastes in hot cells. The high level radioactive solid waste is divided into three main categories, a flammable waste, a non-flammable waste, and a solid reagent waste. A plastic product is categorized into the flammable waste and molten with a heating mantle. The non-flammable waste is cut with a band saw machine for reducing the volume. Among the solid reagent waste, a used adsorbent after the experiments is heated, and an extractant is decomposed for its stabilization. All high level radioactive solid wastes in the hot cells are packed in a high level radioactive solid waste can. The high level radioactive solid waste can is transported to the 2nd High Active Solid Waste Storage in the Tokai Reprocessing Plant in the Japan Atomic Energy Agency.
Abstract: In a nuclear reactor, an array of fuel rods containing stacked uranium dioxide pellets clad with zircalloy is the heat source for a thermodynamic cycle of energy conversion from heat to electricity. After fuel is used in a nuclear reactor, the assemblies are stored underwater in a spent nuclear fuel pool at the nuclear power plant while heat generation and radioactive decay rates decrease before it is placed in packages for dry storage or transportation. A computational model of a Boiling Water Reactor spent fuel assembly is modeled using FLUENT, the computational fluid dynamics package. Heat transfer simulations were performed on the two-dimensional 9x9 spent fuel assembly to predict the maximum cladding temperature for different input to the FLUENT model. Uncertainty quantification is used to predict the heat transfer and the maximum temperature profile inside the assembly.
Abstract: Analysis of the uncertainty quantification related to nuclear safety margins applied to the nuclear reactor is an important concept to prevent future radioactive accidents. The nuclear fuel performance code may involve the tolerance level determined by traditional deterministic models producing acceptable results at burn cycles under 62 GWd/MTU. The behavior of nuclear fuel can simulate applying a series of material properties under irradiation and physics models to calculate the safety limits. In this study, theoretical predictions of nuclear fuel failure under transient conditions investigate extended radiation cycles at 75 GWd/MTU, considering the behavior of fuel rods in light-water reactors under reactivity accident conditions. The fuel pellet can melt due to the quick increase of reactivity during a transient. Large power excursions in the reactor are the subject of interest bringing to a treatment that is known as the Fuchs-Hansen model. The point kinetic neutron equations show similar characteristics of non-linear differential equations. In this investigation, the multivariate logistic regression is employed to a probabilistic forecast of fuel failure. A comparison of computational simulation and experimental results was acceptable. The experiments carried out use the pre-irradiated fuels rods subjected to a rapid energy pulse which exhibits the same behavior during a nuclear accident. The propagation of uncertainty utilizes the Wilk's formulation. The variables chosen as essential to failure prediction were the fuel burnup, the applied peak power, the pulse width, the oxidation layer thickness, and the cladding type.
Abstract: The performances of nuclear fuels and materials are qualified at an irradiation system in research reactors operating under the commercial nuclear power plant conditions. Fuel centerline temperature, coolant temperature, neutron flux, deformations of fuel stack and swelling are important parameters needed to analyze the nuclear fuel performances. The dimensional stability of nuclear fuels is a key parameter measuring the fuel densification and swelling. In this study, the fuel stack elongation is measured using a LVDT. A mockup LVDT instrumented fuel rod is developed. The performances of mockup LVDT instrumented fuel rod is evaluated by experiments.
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: Spacer grid assembly supporting the nuclear fuel rods
is an important concern in the design of structural components of a
Pressurized Water Reactor (PWR). The spacer grid is composed by
springs and dimples which are formed from a strip sheet by means of
blanking and stamping processes. In this paper, the blanking process
and tooling parameters are evaluated by means of a 2D plane-strain
finite element model in order to evaluate the punch load and quality
of the sheared edges of Inconel 718 strips used for nuclear spacer
grids. A 3D finite element model is also proposed to predict the
tooling loads resulting from the stamping process of a preformed
Inconel 718 strip and to analyse the residual stress effects upon the
spring and dimple design geometries of a nuclear spacer grid.
Abstract: Flow coastdown phenomena are very important to secure nuclear fuel integrity during loss of off-site power accidents. In this study, primary coolant flow coastdown phenomena are investigated for the Jordan Research and Training Reactor (JRTR) using a simulation software package, Modular Modeling System (MMS). Two MMS models are built. The first one is a simple model to investigate the characteristics of the primary coolant pump only. The second one is a model for a simulation of the Primary Coolant System (PCS) loop, in which all the detailed design data of the JRTR PCS system are modeled, including the geometrical arrangement data. The same design data for a PCS pump are used for both models. Coastdown curves obtained from the two models are compared to study the PCS loop coolant inertia effect on a flow coastdown. Results showed that the loop coolant inertia effect is found to be small in the JRTR PCS loop, i.e., about one second increases in a coastdown half time required to halve the coolant flow rate. The effects of different flywheel inertia on the flow coastdown are also investigated. It is demonstrated that the coastdown half time increases with the flywheel inertia linearly. The designed coastdown half time is proved to be well above the design requirement for the fuel integrity.
Abstract: The gel-supported precipitation (GSP) process can be
used to make spherical particles (spherules) of nuclear fuel,
particularly for very high temperature reactors (VHTR) and even for
implementing the process called SPHEREPAC. In these different
cases, the main characteristics are the sphericity of the particles to be
manufactured and the control over their grain size. Nonetheless,
depending on the specifications defined for these spherical particles,
the GSP process has intrinsic limits, particularly when fabricating
very small particles. This paper describes the use of secondary
fragmentation (water, water/PVA and uranyl nitrate) on solid
surfaces under varying temperature and vibration conditions to assess
the relevance of using this new technique to manufacture very small
spherical particles by means of a modified GSP process. The
fragmentation mechanisms are monitored and analysed, before the
trends for its subsequent optimised application are described.
Abstract: Energy generated by the force of water in hydropower
can provide a more sustainable, non-polluting alternative to fossil
fuels, along with other renewable sources of energy, such as wind,
solar and tidal power, bio energy and geothermal energy. Small scale
hydroelectricity in Iran is well suited for “off-grid" rural electricity
applications, while other renewable energy sources, such as wind,
solar and biomass, can be beneficially used as fuel for pumping
groundwater for drinking and small scale irrigation in remote rural
areas or small villages. Small Hydro Power plants in Iran have very
low operating and maintenance costs because they consume no fossil
or nuclear fuel and do not involve high temperature processes. The
equipment is relatively simple to operate and maintain. Hydropower
equipment can adjust rapidly to load changes. The extended
equipment life provides significant economic advantages. Some
hydroelectric plants installed 100 years ago still operate reliably. The
Polkolo river is located on Karun basin at southwest of Iran. Situation
and conditions of Polkolo river are evaluated for construction of
small hydropower in this article. The topographical conditions and
the existence of permanent water from springs provide the suitability
to install hydroelectric power plants on the river Polkolo. The
cascade plant consists of 9 power plants connected with each other
and is having the total head as 1100m and discharge about 2.5cubic
meter per second. The annual production of energy is 105.5 million
kwh.
Abstract: The mixed oxide nuclear fuel (MOX) of U and Pu contains several percent of fission products and minor actinides, such as neptunium, americium and curium. It is important to determine accurately the decay heat from Curium isotopes as they contribute significantly in the MOX fuel. This heat generation can cause samples to melt very quickly if excessive quantities of curium are present. In the present paper, we introduce a new approach that can predict the decay heat from curium isotopes. This work is a part of the project funded by King Abdulaziz City of Science and Technology (KASCT), Long-Term Comprehensive National Plan for Science, Technology and Innovations, and take place in King Abdulaziz University (KAU), Saudi Arabia. The approach is based on the numerical solution of coupled linear differential equations that describe decays and buildups of many nuclides to calculate the decay heat produced after shutdown. Results show the consistency and reliability of the approach applied.
Abstract: For gamma radiation detection, assemblies having
scintillation crystals and a photomultiplier tube, also there is a
preamplifier connected to the detector because the signals from
photomultiplier tube are of small amplitude. After pre-amplification
the signals are sent to the amplifier and then to the multichannel
analyser. The multichannel analyser sorts all incoming electrical
signals according to their amplitudes and sorts the detected photons
in channels covering small energy intervals. The energy range of
each channel depends on the gain settings of the multichannel
analyser and the high voltage across the photomultiplier tube. The
exit spectrum data of the two main isotopes studied ,putting data in
biomass program ,process it by Matlab program to get the solid
holdup image (solid spherical nuclear fuel)
Abstract: The presented article deals with the description of a
numerical model of a corridor at a Central Interim Spent Fuel Storage
Facility (hereinafter CISFSF). The model takes into account the
effect of air flows on the temperature of stored waste. The
computational model was implemented in the ANSYS/CFX
programming environment in the form of a CFD task solution, which
was compared with an approximate analytical calculation. The article
includes a categorization of the individual alternatives for the
ventilation of such underground systems. The aim was to evaluate a
ventilation system for a CISFSF with regard to its stability and
capacity to provide sufficient ventilation for the removal of heat
produced by stored casks with spent nuclear fuel.
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.