High-temperature gas-cooled reactors

High-Temperature Gas-Cooled Reactor (HTGR)

The HTGR core consists of a massive pile of hexagonal graphiteblocks, each containing fueled regions, the whole structure being cooledby pressurized helium gas. The graphite serves as the neutron moderator.The fuel itself consists of highly-enriched uranium as the fissile materialand thorium 232 as the fertile. These materials, in the form of thedioxide or carbide, are present as small fuel kernels with ceramic coatings,so that both fuel and its “cladding” are ceramic. The final fuel willbe composed of two particle types, one containing uranium enriched to93 % 235U, the other containing the thorium. As the reactor runs, fissile -233U would build up in the latter particles. The fuel particles are boundinto fuel rods (with graphite as the binder) which are incorporated intothe basic block or element. These elements are stacked. A basic refuelingregion consists of a central stack, which has two vertical control rodchannels, and the six adjacent stacks (without such channels). The centralstack has an additional channel into which boron balls can be pouredas a reserve shutdown system. All of the fuel elements have holesthrough which the coolant flows.

The core and the other NSSS components are contained in a“prestressed-concrete reactor vessel” (PCRV) which has steel-linedcavities for each component. The reactor is fueled through penetrationsabove the central stack of each refueling region; this same penetrationalso serves for the drive for the control rod pair penetrating that stack.There are additional penetrations for piping and removal plugs for servicing(and even replacement) of steam generators, circulators, etc. Thevessel is prestressed with vertical steel tendons and with circumferentialwire wrapping. The PCRV and its contents are extremely massive(100,000,000 Ib); and indeed the core itself is more massive, by aboutone order of magnitude, than the core of an LWR. The PCRV is surroundedby a containment building much like that of a PWR.

The helium gas, at a pressure of 700 psi, is pumped downward throughthe core, and exits with a temperature of 1370 °F. The gas then passes intoone of the 4 or 6 (depending on the plant size) heat exchangers, raisingsteam as usual. Above each steam generator is a circulator which pumps thehelium. The higher reactor operating temperature, which is permitted by thegaseous form of the coolant and the good high-temperature characteristicsof the core, provides steam which can be converted to electrical energy withan efficiency of 39 %, substantially greater than in a water-cooled reactor.Moreover, the potential exists, with development of helium-driven turbogenerators, to raise this efficiency even more.

The accident response of an HTGR is noticeably different than that of awater-cooled reactor, principally because of the mass of the core, were shutdownto occur with complete loss of cooling, no damage to the fuel wouldoccur for about half an hour (as compared with the roughly 1 min of anLWR). The fuel particle, due to its ceramic coating, is much less susceptible

to melting than a metal-clad fuel rod. And, due to the, heat capacity of thecore, it takes much longer to reach elevated temperatures. Moreover, thecore’s structural strength is provided by graphite, whose strength improvesas the temperature rises.To make a complete loss of coolant incredible, flow restrictors areincorporated around PGRV penetrations to reduce helium loss shouldthe vessel integrity be violated there. As a result, helium is always presumedto be in the system. Should all the primary cooling loops becomeunavailable, which is difficult to predict since they are largely independent,the two or three loops of the auxiliary cooling system are activated.These are sized to be sufficient for safe shutdown.

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Обозначение проблемы

Анализ различных мнений по проблеме (С одной стороны…, с другой стороны….)

Выводы (ваше отношение и решение проблемы)

Заключение.

Вариант 7

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Fast Reactors

Uranium resources are not unlimited. Although a gram of 235U“burned” in a reactor, yields more energy than a ton of coal, reasonablyassured U.S. deposits of uranium are only enough to supply, until theyear 2000, reactors as they are presently being built and operated. By

that year there will be roughly 1000 nuclear power plants operating, andthe cumulative uranium requirements for initial cores and for refuelingwill amount to 2 or 3 million tons of U3O8. This is comparable to thetotal amount of uranium that is presently thought to be available at reasonableeconomic and environmental cost. The question is, then afterthe year 2000, how do we supply (1) initial cores for new plants startingup and (2) the continuing fuel needs of the accumulated nuclear powersystem? Even if we agreed not to build any now fission power plants

after that year, fuel requirements for those already built would substantiallyexceed the uranium now thought to be available. The standard solutionto this problem is that late in this century we should begin building“breeder” reactors which not only replace their own fuel, but producean excess to be used in other reactors.

Fast reactors currently use mixed plutonium-uranium fuel producedinitially as a by-product from thermal reactors. The plutonium is sufficientlyconcentrated for the fission reaction to be sustained by fast neutronsso no moderator is necessary. Without a moderator, the core of thereactor is compact and has a high heat rating so very efficient coolantliquidsodium is used to transfer the heat.

In addition, the core can be surrounded by a uranium blanket andthus 'breed' more plutonium.

Doubling time

As we have already mentioned, the doubling time is simply the ratioof the reactor (or fuel cycle) inventory to the amount of excess material.Doubling time is therefore proportional to the fissile inventory dividedby the breeding gain.

Strictly speaking, this definition would be correct only if the excessfissile material were held in storage until a sufficient amount accumulatedto supply the entire fissile inventory of a new reactor. In the more likelysituation, excess production of a number of reactors would contribute to asingle new reactor, so that even the excess fissile material would soon beproducing its own excess. In this situation, the rate of excess productionhas, effectively, to be compounded to yield a real doubling time.

In other respects, too we have not defined doubling time preciselyenough. For example, we have not given any attention to how differinginput and output isotopic mixes affect the definition of doubling time.However, for present purposes, the most important clarification is to distinguishthe doubling time associated with only the fuel inventory of thereactor itself and that associated with the fuel, not only in the reactor, butalso in the fuel cycle, i.e., in storage, reprocessing, fabrication, so on. Thepractical doubling time of a breeder reactor system is the time it takes thereactor and its facilities to yield enough excess fissile material to supplyan additional reactor and its support facilities.

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3. Составьте аннотацию к тексту (2-3 предложения).

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5. Выскажите свое мнение по проблемам, изложенным в тексте:

Обозначение проблемы

Анализ различных мнений по проблеме (С одной стороны…, с другой стороны….)

Выводы (ваше отношение и решение проблемы)

Заключение.

Вариант 8

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