USE OF NUCLEAR REACTORS FOR HEAT SUPPLY

A. V. Kuzmin, S. A. Belyaev

(Presented by the scientific seminar of the Department of KKU)

In recent years the scale and pace of construction of nuclear power installations in the USSR and abroad have increased substantially. New generating capacities are being introduced, and in the immediate future will mainly be introduced, through the construction of condensing nuclear power plants (NPPs) [1]. In the future, NPPs will gradually displace stations burning organic fuel. The increasing use of nuclear fuel in power engineering does not solve the problem of fuel shortage [2], since the effect of nuclear fuel on the overall fuel-and-energy balance will be relatively small ($\sim 25\%$). The enormous expenditures of high-quality organic fuel for the production of low- and medium-potential heat for industrial and municipal heat consumption raise the question of the earliest possible use both of the combined method of producing electric power and heat in nuclear power engineering, and of nuclear district boiler plants (NDBP) for centralized heat supply. There are not very many examples of district-heating NPPs, and there are no NDBPs at all. The combined method of producing electric power and heat has become widespread mainly in small-scale nuclear power engineering [3]. The first assessment of the possibility of using nuclear reactors for district heating and heat supply was made in [4].

The various processes associated with the consumption of heat without its further conversion into other forms of energy are divided into two categories:

a) heat used to meet municipal and domestic needs;
b) heat used for technological needs. The scale of production of low-potential heat is comparable with the total capacity of all large power stations of the country. At present, municipal and domestic needs account for about 70%, and technological needs for about 30%, of the country’s total heat consumption. This ratio will remain the same in the near future [5].

For municipal and domestic needs, heat is consumed at low coolant temperatures ($\sim 150^\circ\mathrm{C}$); the load schedule is characterized by high stability. This creates attractive possibilities for designing elements of the nuclear installation of an NDBP from inexpensive materials, simple in design, for regulating and maintaining power. In the NDBP scheme, as the heat generator there may be used

any reactor types developed in the USSR and abroad were used. We evaluated the operation of a gas-graphite reactor, a reactor with a heavy-water moderator and coolant (of the “Adam” type), boiling and water-water reactors with separate circuits and in a monoblock arrangement (of the ABV-1, 5 type [3]). All the variants listed were compared at an ARK thermal capacity of 100 Gcal/h (some reactors were also compared at a capacity of 300 Gcal/h).

The annual load-duration curve shown in Fig. 1 was compiled for the eastern regions of the country with \(t_p^n = -40^\circ\mathrm{C}\). In this case the utilization of the ARK in the base part of the curve is about 6000 hours. It is obvious that for regions with a lower design outdoor-air temperature, the utilization of the ARK in the base part of the load curve will increase and short-term load variations will be reduced to a minimum. The peak part of the load is assumed to be covered by a peak hot-water boiler house operating on fuel oil or gas. The water at the inlet to the heating network after the ARK has a maximum temperature of \(110^\circ\mathrm{C}\), and in the return main, \(30\text{–}70^\circ\mathrm{C}\) (Fig. 2).

Fig. 1. Annual heat-load curve

Fig. 2. Temperature chart of the heating network

When designing an ARK it is necessary to take into account the following basic requirements:

a) the ARK must have a high degree of radiation safety; the ARK design must reliably protect the heating networks from radioactive contamination;

b) it is necessary to strive for the maximum possible proximity of the ARK to the center of the heat loads;

c) the heat generator—the reactor—must possess high operational reliability, simplicity of control, and a wide range of power regulation;

d) the ARK must be loaded in the base part of the load curve, since only in this case can capital investments, significantly greater than those for ordinary boiler houses, be justified;

e) it is desirable to strive to increase the reactor campaign as much as possible in order to simplify ARK maintenance to the maximum;

f) construction and installation work must be reduced to a minimum.

The first three of the above requirements make it necessary to select as the heat source reliable reactors that have been mastered in production and operation. Heavy-water reactors, despite their physical advantages, have a very bulky network of auxiliary systems, which complicates operation [6].

Data on the design of boiling reactors at low parameters and power has not been encountered by us in the literature. The reliability of reactors of this type is problematic.

The most preferable from the standpoint of radiation safety and bringing the nuclear installation closer to the center of heat loads are gas-graphite reactors and the water-water reactors well mastered in our country and abroad. The gaseous coolant (CO₂) has weak neutron absorption, as a result of which its leakage from the reactor leads to only a very slight increase in reactivity. This is a characteristic property ensuring the safety of operation of these reactors. Gas-graphite reactors are characterized by simplicity of design, ease and convenience of operation. In the range of the required, rather low temperatures up to 400°C, carbon dioxide does not corrode the structural materials of the primary circuit. The disadvantages of these reactors are the large dimensions of the heat-exchange apparatus and the active zone, as well as the need for a large amount of purchased electric power required for pumping the gas in the primary circuit.

Water-water reactors possess the property of self-regulation, used for automatic regulation during transient processes, which makes it possible not to use fast-acting expensive control systems on them. Low parameters \((P = 10—20\ \text{ata},\ t < t_s)\), and the use of natural circulation also significantly simplify the auxiliary systems of the primary circuit, facilitating servicing and control of the reactor. All these advantages of the reactor make it possible to consider it the most promising. A possible scheme of an APH with a water-water reactor, made according to a separated arrangement, is given in Fig. 3.

Fig. 3. Schematic heat diagram of an APH: 1 — reactor; 2 — intermediate heat exchanger; 3 — network preheater; 4 — network pump; 5 — recirculation pump; 6 — chemical water-treatment pump; 7 — heat exchanger; 8 — deaerator; 9 — vapor cooler; 10 — raw water; 11 — tank; 12 — storage tank; 13 — make-up pump of the heat network

The requirements of ensuring normal operating conditions and bringing the heat source closer to consumers also compel special attention, in designing APHs, to the choice of site and to measures and engineering means for ensuring radiation safety. In the practice of evaluating sites for nuclear installations, previously the basic principle for solving the safety problem was easily discerned: removing the reactor from a populated area. The U.S. Atomic Energy Commission, in the second edition of the “Interim Design Rules” (1962), reaches an interesting conclusion: removing nuclear power plant reactors from places of consumption is a rather ineffective means of increasing safety [7].

The basic criteria for evaluating a site according to the USAEC are reduced to the calculated determination of three zones around the reactor site: the exclusion zone, the low-population zone, and the distance to a large populated area. The dimensions of these zones are affected by the size of the accident; therefore, in calculations the following potentially conceivable conditions of the limiting possible accident are allowed: a large release of activity, poor ventilation of the near-ground

layer of air and, most importantly, the destruction of protective equipment. The AEC believes that if the probability of these events were zero, or if there were experience in operating protective equipment under emergency conditions, there would be no need to introduce the second and third zones.

The data of theoretical calculations for determining the radii of protective zones are given in Table 1. It should be noted that these data were obtained for nuclear power reactors operating at higher

Table 1

Radii of exclusion zones for reactors
of various power levels according to U.S. AEC rules

parameters than APC reactors. The higher reliability of the protective equipment at APC reactors can probably be achieved with much less difficulty. This gives grounds for believing that construction of APC reactors will be possible in large populated localities. Of course, this will pose many problems in the design of auxiliary systems, engineering means of radiation safety and protective enclosures, layout of the boiler-house building, improvement of site assessment, etc., etc.; but the obvious advantages of APC reactors over ordinary boiler plants (a higher efficiency, less air pollution, a lower cost of the heat supplied (Table 2) owing to lower production costs) require their rapid solution.

Table 2

Cost of heat obtained by various methods

The use of nuclear fuel for generating low-potential heat is undoubtedly a promising direction for improving the country’s fuel-and-energy balance.

References

1. G. V. Ermakov. Development of nuclear power engineering in the USSR. Teploenergetika, 1969, No. 6.

2. A. P. Aleksandrov. Nuclear power engineering and its role in technological progress. Atomnaya energiya. Vol. 25, issue 5, 1968.

3. V. A. Zhilytsov et al. Application of water-water reactors in small-scale nuclear power engineering. Atomnaya energiya. Vol. 26, issue 5, 1969.

4. A. S. Gorshkov et al. Use of nuclear reactors for district heating and heat supply. Report 319, presented by the USSR to the Third International Conference on the Peaceful Uses of Atomic Energy. Geneva, 1964.

5. L. A. Melent’ev, M. A. Styrikovich, E. O. Shteingauz. The Fuel and Energy Balance of the USSR (Principal Questions of Economics and Planning). Gosenergoizdat, 1962.

6. I. Vives­tad, S. Mileikovskii. Proceedings of the Second International Conference on the Peaceful Uses of Atomic Energy. Selected Papers of Foreign Scientists. Vol. 4. Moscow, 1959.

7. Yu. V. Sivintsev. Safety of nuclear reactors. Atomnaya tekhnika za rubezhom, 1966, No. 10.

8. V. A. Orlov, G. N. Leus. In the collection: Problems of the North. Issue 10. Moscow, Nauka, 1964, p. 207.

9. A. Ya. Avrukh. Problems of the Cost of Electric and Thermal Energy. Moscow, Gosenergoizdat, 1963.