Process Heat Transfer - By D. Q. Kern - Free ebook download as PDF File .pdf), Text File .txt) or read book online for free. HE,Evaporator, heat transfer. PROCESS HEAT TRANSFER. BY. DONALD Q. KERN. D. Q. Kern Associates, and. Professorial Lecturer in Chemical Engineering. Case Institute of Techowlogy . A heat transfer textbook / John H. Lienhard IV and. John H. Lienhard V — 3rd ed. — Cambridge, MA: J.H. Lienhard V, c Includes bibliographic references.
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PROCESS HEAT TRANSFER. BY. DONALD Q. KERN. D. Q. Kern Associates, and. Professorial Lecturer in Chemical Engineering. Case Instiiute of Technology . PROCESS HEAT TRANSFER. BY. DONALD Q. KERN. D. Q. Kern Associates, and. Professorial Lecturer in Chemical Engineering. Case Institute of Technology . Process Heat Transfer - D.Q. Kern - Ebook download as PDF File .pdf) or read book online.
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The reactor's void coefficient depends on the core content; it ranges from negative with all the initial absorbers to positive when they are all removed.
The moisture and temperature of the outlet gas is monitored; an increase of them is an indicator of a coolant leak. The reactor has two independent cooling circuits, each having four main circulating pumps three operating, one standby.
The cooling water is fed to the reactor through lower water lines to a common pressure header one for each cooling circuit , which is split to 22 group distribution headers, each feeding 38—41 pressure channels through the core, where the feedwater boils. The mixture of steam and water is led by the upper steam lines, one for each pressure channel, from the reactor top to the steam separators , pairs of thick horizontal drums located in side compartments above the reactor top; each has 2.
The resulting feedwater is led to the steam separators by feedwater pumps and mixed with water from them at their outlets.
From the bottom of the steam separators, the feedwater is led by 12 downpipes from each separator to the suction headers of the main circulation pumps, and back into the reactor. The turbine consists of one high-pressure rotor and four low-pressure ones.
Five low-pressure separators-preheaters are used to heat steam with fresh steam before being fed to the next stage of the turbine. The uncondensed steam is fed into a condenser, mixed with condensate from the separators, fed by the first-stage condensate pump to a chemical purifier, then by a second-stage condensate pump to four deaerators where dissolved and entrained gases are removed; deaerators also serve as storage tanks for feedwater.
From the deaerators, the water is pumped through filters and into the bottom parts of the steam separator drums. Each pump has a flow control valve and a backflow preventing check valve on the outlet, and shutoff valves on both inlet and outlet. Each of the pressure channels in the core has its own flow control valve so that the temperature distribution in the reactor core can be optimized.
Each channel has a ball type flow meter. With few absorbers in the reactor core, such as during the Chernobyl accident, the positive void coefficient of the reactor makes the reactor very sensitive to the feedwater temperature. Bubbles of boiling water lead to increased power, which in turn increases the formation of bubbles. After absorbers were introduced in the fuel assembly, permanently assuring a negative void coefficient at the cost of higher enrichment requirements of the uranium fuel.
If the coolant temperature is too close to its boiling point, cavitation can occur in the pumps and their operation can become erratic or even stop entirely. At low reactor power, therefore, the inlet temperature may become dangerously high. The water is kept below the saturation temperature to prevent film boiling and the associated drop in heat transfer rate. The reactor is tripped in cases of high or low water level in the steam separators with two selectable low-level thresholds ; high steam pressure; low feedwater flow; loss of two main coolant pumps on either side.
These trips can be manually disabled. The level of water in the steam separators, the percentage of steam in the reactor pressure tubes, the level at which the water begins to boil in the reactor core, the neutron flux and power distribution in the reactor, and the feedwater flow through the core have to be carefully controlled. The level of water in the steam separator is mainly controlled by the feedwater supply, with the deaerator tanks serving as a water reservoir. The reactor is equipped with an emergency core cooling system ECCS , consisting of dedicated water reserve tank, hydraulic accumulators, and pumps.
ECCS piping is integrated with the normal reactor cooling system. In case of total loss of power, the ECCS pumps are supposed to be powered by the rotational momentum of the turbogenerator rotor for the time before the diesel generators come online.
The Chernobyl disaster occurred during a botched test of this system. The ECCS has three systems, connected to the coolant system headers. In case of damage, the first ECCS subsystem provides cooling for up to seconds to the damaged half of the coolant circuit the other half is cooled by the main circulation pumps , and the other two subsystems then handle long-term cooling of the reactor.
The third group is a set of electrical pumps drawing water from the deaerators. The short-term pumps can be powered by the spindown of the main turbogenerators. ECCS for long-term cooling of the damaged circuit consists of three pairs of electrical pumps, drawing water from the pressure suppression pools; the water is cooled by the plant service water by means of heat exchangers in the suction lines.
Each pair is able to supply half of the maximum coolant flow. ECCS for long-term cooling of the intact circuit consists of three separate pumps drawing water from the condensate storage tanks, each able to supply half of the maximum flow. Some valves that require uninterrupted power are also backed up by batteries. The distribution of power density in the reactor is measured by ionization chambers located inside and outside the core.
The physical power density distribution control system PPDDCS has sensors inside the core; the reactor control and protection system RCPS uses sensors in the core and in the lateral biological shield tank. The external sensors in the tank are located around the reactor middle plane, therefore do not indicate axial power distribution nor information about the power in the central part of the core.
There are over radial and 12 axial power distribution monitors, employing self-powered detectors.
Reactivity meters and removable startup chambers are used for monitoring of reactor startup. Total reactor power is recorded as the sum of the currents of the lateral ionization chambers. The moisture and temperature of the gas circulating in the channels is monitored by the pressure tube integrity monitoring system.
The RCPS system consists of movable control rods. Both systems, however, have deficiencies, most noticeably at low reactor power levels. Below those levels, the automatic systems are disabled and the in-core sensors are not accessible.
Without the automatic systems and relying only on the lateral ionization chambers, control of the reactor becomes very difficult; the operators do not have sufficient data to control the reactor reliably and have to rely on their intuition.
During startup of a reactor with a poison-free core this lack of information can be manageable because the reactor behaves predictably, but a non-uniformly poisoned core can cause large nonhomogenities of power distribution, with potentially catastrophic results.
The reactor emergency protection system EPS was designed to shut down the reactor when its operational parameters are exceeded. However, the slow insertion speed of the control rods, together with their design causing localized positive reactivity as the displacer moves through the lower part of the core, created a number of possible situations where initiation of the EPS could itself cause or aggravate a reactor runaway.
The computer system for calculation of the reactivity margin was collecting data from about 4, sources.
Its purpose was to assist the operator with steady-state control of the reactor. Ten to fifteen minutes were required to cycle through all the measurements and calculate the results. The operators could disable some safety systems, reset or suppress some alarm signals, and bypass automatic scram , by attaching patch cables to accessible terminals.
This practice was allowed under some circumstances. The reactor is equipped with a fuel rod leak detector. A scintillation counter detector, sensitive to energies of short-lived fission products, is mounted on a special dolly and moved over the outlets of the fuel channels, issuing an alert if increased radioactivity is detected in the steam-water flow. The RBMK design was built primarily to be powerful, quick to build and easy to maintain.
Full physical containment structures for each reactor would have more than doubled the cost and construction time of each plant, and since the design had been certified by the Soviet nuclear science ministry as inherently safe when operated within established parameters the Soviet authorities assumed proper adherence to doctrine by workers would make any accident impossible.
Additionally, RBMK reactors were designed to allow fuel rods to be changed without shutting down as in the pressurized heavy water CANDU reactor , both for refueling and for plutonium production for nuclear weapons. This required large cranes above the core.
In the Chernobyl accident , the pressure rose to levels high enough to blow the top off the reactor, breaking open the fuel channels in the process and starting a massive fire when air contacted the superheated graphite core.
After the Chernobyl accident, some RBMK reactors were retrofitted with a partial containment structure in lieu of a full containment building , which surround the fuel channels with water jackets in order to capture any radioactive particles released. The bottom part of the reactor is enclosed in a watertight compartment.
There is a space between the reactor bottom and the floor. In the event of an accident, which was predicted to be at most a rupture of one or two pressure channels, the steam was to be bubbled through the water and condensed there, reducing the overpressure in the leaktight compartment.
The flow capacity of the pipes to the pools limited the protection capacity to simultaneous rupture of two pressure channels; a higher number of failures would cause pressure buildup sufficient to lift the cover plate "Structure E", after the explosion nicknamed "Elena" , sever the rest of the fuel channels, destroy the control rod insertion system, and potentially also withdraw control rods from the core.
The leaktight compartments around the pumps can withstand overpressure of 0. The distribution headers and inlets enclosures can handle 0. The reactor cavity can handle overpressure of 0. The pressure suppression system can handle a failure of one reactor channel, a pump pressure header, or a distribution header. Leaks in the steam piping and separators are not handled, except for maintaining slightly lower pressure in the riser pipe gallery and the steam drum compartment than in the reactor hall.
These spaces are also not designed to withstand overpressure. The steam distribution corridor contains surface condensers. The fire sprinkler systems , operating during both accident and normal operation, are fed from the pressure suppression pools through heat exchangers cooled by the plant service water, and cool the air above the pools.
Jet coolers are located in the topmost parts of the compartments; their role is to cool the air and remove the steam and radioactive aerosol particles.
The air removal is stopped automatically in case of a coolant leak and has to be reinstated manually. For the nuclear systems described here, the Chernobyl Nuclear Power Plant is used as the example. The generators are connected to their common transformer by two switches in series. Between them, the unit transformers are connected to supply power to the power plant's own systems; each generator can therefore be connected to the unit transformer to power the plant, or to the unit transformer and the generator transformer to also feed power to the grid.
In case of total external power loss, the essential systems can be powered by diesel generators. The 7A, 7B, and 8B boards are also connected to the three essential power lines namely for the coolant pumps , each also having its own diesel generator.
In case of a coolant circuit failure with simultaneous loss of external power, the essential power can be supplied by the spinning down turbogenerators for about 45—50 seconds, during which time the diesel generators should start up.
The generators are started automatically within 15 seconds at loss of off-site power. The generator's stator is cooled by water while its rotor is cooled by hydrogen.
The hydrogen for the generators is manufactured on-site by electrolysis. The Chernobyl plant was equipped with both types of turbines; Block 4 had the newer ones.
It was designed and constructed with several design characteristics that proved dangerously unstable when operated outside their design specifications. The decision to use a superheated, vacuum-isolated graphite core with natural uranium fuel allowed for massive power generation at only a quarter of the expense of heavy water reactors, which were more maintenance-intensive and required large volumes of expensive heavy water for startup.
However, it also had unexpected negative consequences that would not reveal themselves fully until the Chernobyl disaster. Light water the ordinary H 2 O is both a neutron moderator and a neutron absorber.
This means that not only can it slow down neutrons to velocities in equilibrium with surrounding molecules "thermalize" them and turn them into low-energy neutrons that are far more likely to interact with the uranium nuclei than the fast neutrons produced by fission initially , but it can also absorb some of them outright.
Heavy water is also a good neutron moderator, but is expensive to produce and does not absorb neutrons as easily, so the use of enriched fuel is not required to produce a meaningful power output. In RBMKs, light water was used as a coolant; moderation was mainly carried out by graphite. As graphite already moderated neutrons, light water had a lesser effect in slowing them down, but could still absorb them.
This means that the reactor's reactivity adjustable by appropriate neutron-absorbing rods had to account for the neutrons absorbed by light water. Because of this lower density of mass, and consequently of atom nuclei able to absorb neutrons , light water's neutron-absorption capability practically disappears when it boils.
This allows more neutrons to fission more U nuclei and thereby increase the reactor power, which leads to higher temperatures that boil even more water, creating a thermal feedback loop. In RBMKs, generation of steam in the coolant water would then in practice create a void, a bubble that does not absorb neutrons; the reduction in moderation by light water is irrelevant, as graphite is still moderating the neutrons. However, the loss of absorption would dramatically alter the balance of neutron production, causing a runaway condition in which more and more neutrons are produced, and their density grows exponentially fast.
Such a condition is called a positive void coefficient , and the RBMK has the highest positive void coefficient of any commercial reactor ever designed. It should be noted that a high void coefficient does not necessarily make a reactor inherently unsafe , as some of the fission neutrons are emitted with a delay of seconds or even minutes post-fission neutron emission from daughter nuclei , so steps can be taken to reduce the fission rate before it gets too high.
However, it does make it considerably harder to control the reactor especially at low power and makes it imperative that the control systems are very reliable and the control room personnel regardless of rank or position are rigorously trained in the peculiarities and limits of the system.
Neither of these requirements were in place at Chernobyl: Some later RBMK designs did include control rods on electromagnetic grapples, thus controlling the reaction speed and, if necessary, stopping the reaction completely.
This new number decreases the possibility of a low-coolant meltdown. Following Legasov's death,  all remaining RBMKs were retrofitted with a number of updates for safety. The largest of these updates fixes the RBMK control rod design.
The control rods have graphite tips attached, which prevent coolant water from entering the space vacated as the rods are withdrawn. In the original design, those displacers, being shorter than the height of the core, left columns of water at the bottom when the rods were fully extracted; during insertion, the graphite would first displace that water, locally increasing reactivity. Also, when the rods were in their uppermost position, the absorber tips were outside the core, requiring a relatively large displacement before achieving a significant reduction in reactivity.
These design flaws were likely the final trigger of the first explosion of the Chernobyl accident, causing the lower part of the core to become supercritical when they tried to shut down the highly destabilized reactor by reinserting the rods. In addition, RELAPD models of RBMK reactors were developed for use in integrated thermal-hydraulics-neutronics calculations for the analysis of specific transients in which the neutronic response of the core is important.
From May to December , Leningrad -1 was offline while repairs were made related to deformed graphite moderator blocks. The month project included research and the development of maintenance machines and monitoring systems.
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