by Acad. Nikolai PONOMAREV-STEPNOI; Nikolai KUKHARKIN, Cand. Sc. (Tech.); Vadim GREBENNIK, Senior Researcher; National Research Center "Kurchatov Institute", Moscow, Russia

For many years world powers have been involved with creating high-temperature helium-cooled reactors (HTHR) using helium as a heat transfer medium (coolant). Their essential feature and advantage is in a very high operating temperature, up to 1,000 °C. Such reactors make it possible to gain a high efficiency factor in the generation of electrical energy. More than that, they are effective in heat supply for technological processes in the petroleum, chemical, metallurgical and other industries. They produce hydrogen that allows to save natural fuel and cut environmental pollution. Such reactors expand nuclear energy uses and win new avenues for consumption.

A GLANCE INTO HISTORY

The first steps in designing gas-cooled reactors were made in the United States, Federal Germany and some other countries back in the 1950s. Initially Western countries viewed HTHR chiefly as electrical power generators, for such reactors offered certain advantages, namely high efficiency (40 percent), lower heat discharges and lower consumption of water coolant. They ensured an economical burn-out cycle, effective air heat removal, and high safety.

High-temperature gas-cooled reactor in shorthand.

HTHR-related technologies made certain headway in the 1960s, when small experimental reactors were developed: "Dragon" (Britain), "Peach Bottom" (USA), and AVR (Germany). "Dragon" and "Peach Bottom" stayed in service for about 10 years, and AVR--for as long as twenty, showing good reliability. Experts gained a lot of experience in running such setups, and carried out a great amount of helium-related research.

In the latter half of the 1970s the United States commissioned a reactor at the "Fort St. Vrain" (FSV) nuclear power station; its output power was 330 MW (electricity). Germany put into service a demonstration thorium high-temperature reactor (THTR), its output power was 300 MW (electricity). These setups were in service up until the mid-1980s.

The scope and structure of fuel consumption in the power supply of some countries promised a big potential market for HTHR to replace petroleum and gas. Therefore from the late 1970s in the focus was on a combined output of electrical and fuel energy of medium and high potential in areas where the advantages of such reactors were obvious: they provide for parsimony in using high-calory fuels like gas and petroleum, and guarantee a higher substitution of fossil fuels per unit of output power.

These considerations prodded the United States to begin work on large-scale projects like "Fulton" and "Summit" with a rated heat capacity of 860 and 1,160 MW, respectively. Such setups were meant for use in the petroleum and other industries for combined out-

put electric and high-potential thermal power. Orders for eight nuclear power generators were placed with nuclear hardware manufacturers on the home market; however, the crisis of 1974 compelled the US government to truncate this program, HTHR including.

In spite of the licensing and technical snags, late in 1981 the United States brought the FSV reactor to power, thus proving its design characteristics and operational safety. Thereupon a number of power firms showed an interest in a major nuclear generating network, HTHR inclusive, with a rated capacity of 2,240 MW (heat output power) for electricity and steam generation. Labeled HTGR-SC/C, this project was the first newborn baby in the generation of commercial nuclear reactors with a helium coolant and a moderate temperature of the heat-transfer agent (coolant) of 750 °C.

Next came the HTGR-R setup, a more laborious project meant to cater to heat supplies; it was followed by HTGR-GT. The potential market of HTHR-related technology was estimated in the United States at 500 GW (heat output power) for a period up until 2020, according to the US Energy Department.

In Germany, the prototype THTR-300 reactor equipped with spherical fuel elements* confirmed the

* Fuel element, a central reactor unit containing a fissile material (fisser) and ensuring a reliable heat removal from the fuel to the coolant.--Ed.

design characteristics in voltage within its ferroconcrete pressure vessel and demonstrated its high her-meticity to helium. Thereafter a program was adopted for a broader commercial use of this type reactors upon their technical assimilation. This innovative trend was further adopted in Japan, China, the Republic of South Africa, and other countries.

In our country active research in this area began in the 1960s with the designing of the experimental ABTU-15 reactor and the pilot production plant ABTU-c-50 for power generation and radiative modification of various materials (polyethylene, wood, rubber and the like). This work was initiated and supervised by the Kurchatov Atomic Energy Institute (the National Research Center "Kurchatov Institute" today); the Moscow branch of the Central Boiler and Turbine Institute (today the All-Russia Research and Design Institute of Atomic Engineering) became the chief designer of the project.

Other organizations joined later, in the early 1970s, for one, the Experimental Design Office of Machine-Building in Gorky (now Nizhni Novgorod), which took charge of the HTHR designing job. Under the scientific guidance of "Kurchatovites" it created the VG-400, VGM, and VGM-P setups for generation of electrical and high-potential heat energy. Simultaneously, an experimental and technological base was built. In

Fuel for the reactor's core.

1978 the Kurchatov Institute commissioned a major reactor loop*, PG-100, for upgrading the technology and resource tests of spherical fuel elements at the Materials Science Reactor MR. Problems related to reactor physics were being tackled at "Astra" and GROG stands.

The United States, Japan, Federal Germany and other West European countries also had a good experimental potential. In Japan, the large HENDEL stand with an aggregate capacity of its heaters equal to 16 MW (heat) became a unique base for testing the performance of high-temperature heat exchangers, steam generators, heat insulation and fuel assemblies, and other facilities.

Research scientists at home and abroad discovered certain important advantages of HTHR compared with light-water reactors**, namely higher electrical power generation efficiency (as high as 50 percent in the direct gas-turbine cycle); heat used in technological production (hydrogen output, ammonia synthesis); using uranium-, plutonium- and thorium-based fuels, and other assets. Add higher safety (self-shielding,

* Reactor (in-pile) loop, an independent circulation loop designed for experimental uses; it contains one or several channels.--Ed. ** Light-water reactor, a nuclear generator in which common (light) water is used both as moderator and coolant. There are two kinds of such reactors: those with pressurized (1) and boiling (2) water.--Ed.

lower risk of fuel core melting in major accidents), and the burning of long-lived actinides to scale down radioactive effects on the environment. These and other merits allowed HTHR to carve out a niche in the structure of energy supply of the future.

REACTOR'S PHYSICAL AND TECHNICAL CHARACTERISTICS

In choosing the heat transfer agent (coolant) we should take account of a variety of characteristics: thermal-physical (density, heat capacity, viscosity, heat conductivity); nuclear-physical (action on criticality, radiation resistance, activation); chemical (compatibility with structural materials); and technological ones (toxicity, heat stability, fluidity, explosion and fire hazard, cost, accessibility). Since the main designation of HTHR is to obtain high-temperature (up to 1,000 °C) heat energy, helium was chosen as a gaseous heat transfer agent. The fuel core was to be formed by spherical or prismatical fuel elements composed of microfuel (with 500 µm UO₂ and UC cores, and shielding layers of pyrocarbon and Si carbide) within a graphite matrix.

Gases other than helium were turned town for a number of reasons. Thus hydrogen, for all its good thermal-physical properties, poses a fire risk when

mixing with air; in addition, it exhibits high chemical activity on contacting structural materials at 800 °C and higher. Nitrogen was no good because of its inferior thermal-physical characteristics and effect on radioactivity. Carbon dioxide (CO₂) was not desirable either: at high temperature values it dissociates (breaks apart), with its dissociation products interacting vigorously with graphite, a process resulting in the mass transfer of carbon to the cold regions of the loop.

In the long run all HTHR power plants--those in service, under construction or at the gestation stage--use helium, actually the only coolant complying with most standards for this type reactor. Owing to the chemical inertness of helium, the nuclear fuel and the structural materials of the core can work at high temperature. In practical terms this gas does not absorb and dissipate neutrons, and is not activated by radiation. Although in its specific fuel capacity and the amount of power expended for pumping, helium is second to hydrogen and carbon dioxide, it has good heat conductivity even at moderate pressure (40-50 kgs/cm²), and ensures excellent conditions for the removal and transfer of heat energy in the primary circuit.* This allows to achieve a higher energy intensity in the core and contract the surface of the heat-exchange equipment-compared, say, with CO₂.

It was believed up until recently that since helium exhibits higher fluidity at high temperatures, the work of helium-moderated reactors was wasteful financially, what with the high cost of this gas and extra operational expenses. It was found, however, that at temperatures below 800 °C and pressure under 6 MPa helium does not diffuse through steel. Its diffusion through pipes detected in experiments was the result of submicro-scopical defects in metal appearing at temperatures above 600 °C and a sufficiently high pressure. As proved experimentally, this problem of containing helium in the circuit is solved with much success given the high quality of welding, equipment and its assembly.

According to foreign estimates based on operational data, the cost of helium is less than 1 percent of the cost of electricity generated at major nuclear stations equipped with HTHR, and its losses are largely due to its process extraction (sampling).

The performance of the reactor loop PG-100 argues in favor of the new type setups. This is a fairly big stand 500 m large, and provided with 300 fittings. Its power conduits are 3,500 meters long. The losses of helium in it, including control samplings, do not exceed 0.3 per-

* Reactor primary circuit, a system responsible for a coolant's circulation and heat removal from the fuel core.--Ed.

cent per 24 hours. Good results were also obtained at AVR in Germany (1977): even when the temperature of the coolant was raised from 750 to 950 °C, no substantial losses of the gas were registered.

It would be in place to stress this point: along with helium there is also graphite used in HTHR as moderator, reflector and major structural material in the fuel core, and this under rather rigorous heat conditions. Between 1978 and 1990 the Kurchatov Center, the Research Institute of Nuclear Reactors (Dimitrovgrad, Ulyanovsk Region) and the NIIGraphite Research Center in Moscow carried out a series of tests to evaluate the radiation resistance of graphites (GR-180, GRP-2) used in domestic reactor engineering. They were tested in a wide temperature (300-1,200 °C) and neutron fluence* (2 • 10²² n/cm) range, critical values including. The above fluence value was used as a criterion of the serviceability of the material irradiated in a free unloaded condition as it moved into the stage of intensive swelling resulting in a dramatic worsening of mechanical and thermal-physical characteristics.

These tests as well as bits of evidence obtained abroad made it possible to plot a generalized curve for ultimate serviceability of the material used. The conclusion was that graphites manufactured according to conventional electrode technology did not comply with rated resource characteristics. Attempts to upgrade these indicators through optimization of the granulometric composition of the filler and by increasing volume density (such attempts were made by the NIIGraphite Research Center in the early 1980s) did not yield positive results.

Simultaneously, tests were made on model materials with the use of special techniques. One such material, MPG-6, was employed in the electronic industry. Manufactured on the basis of uncalcinated petroleum coke, it displayed essential structural and other differences compared with conventional graphites (counting in differences in initial characteristics as well). Moreover, it exhibited higher radiation stability, at high temperatures in particular.

Model physical experiments argued in favor of the mainstream trend in HTHR graphites, and that was creating a monophasic material with minimal differences of crystallites (crystalline grains) in dimensionality. In the 1980s NIIGraphite got down to this job by using a composite filler based on uncalcinated coke.

Several technological versions of the material dubbed GR-1 were tested under laboratory conditions at the

* Neutron fluence (fluence of neutrons), a value equal to the ratio of neutrons incident in a given period of time on a surface perpendicular to the neutron flux and the area of this surface.--Ed.

Self-smothering of the reactor through negative temperature and power feedback.

Kurchatov and Dimitrovgrad research centers. The data on dimension stability and on changes of characteristics in the 600 to 1,200 °C temperature range to the fluence value of 2 • 10²² n/cm² showed the high radiation resistance of GR-1 compared with conventional materials. But irradiation experiments had to be continued up to the critical fluence values--this was particularly true of GR-1 production batches. The technology of GR-1 production was developed by the NIIGraphite Research Center and its partner, the plant involved with the manufacture of graphite items at Vyazma, Smolensk administrative region. Also, large-scale investigations were needed into the physical-mechanical, radiation and corrosion characteristics of graphite. One had to compute and validate the serviceability of graphite materials as well. Performance tests were likewise needed at home research reactors, what with a wide temperature range and stringent radiation conditions.

Thus, the favorable combination of the helium coolant and graphite underlies the essential advantages of HTHR, that is the good neutron-physical characteristics at high temperatures. The use of uranium-graphite fuel elements and the helium heat transfer agent is a good economy of neutrons. In addition, HTHR setups can operate in different fuel cycles involving uranium, plutonium and thorium as well.

SAFETY FACTOR

A complex of technical procedures and intrinsic characteristics ensures the safety of high-temperature gas-cooled reactors. The point is that their fuel core is composed largely of graphite having a sublimation temperature of 3,600 °C, which excludes its melting. There occurs no dramatic temperature jump with the loss of the coolant thanks to the high heat capacity of the fuel core. In addition, HTHR possess a high negative power coefficient of reactivity.* This is a major security factor preventing spontaneous power surges. An inert gas--helium in our case--rules out chemical reactions with the fuel and structural materials. Furthermore, helium is not activated: if used as a coolant, no problems of phase transitions. Such power generators can be placed near housing estates and industrial enterprises, and thus keep down losses during heat power transportation, at high temperature in particular.

* Power coefficient of reactivity, a value used for assessing the effect of the power of a reactor on its reactivity. Negative values have a positive effect on its self-shielding: with a power jump or a lower flow rate of the coolant, the reactor will get self-smothered. Also, a negative value of the power coefficient boosts its neutron and thermohydraulic stability.--Ed.