This work is devoted to finding solutions to technical problems so as to improve the technical and economic

performance of NPPs. The possibilities of using neutron moderation energy in gas-cooled water-moderated

channel reactors and triple steam superheating in a Rankine cycle are examined. It is shown that a high

degree of conversion into mechanical energy is attained, well-developed and widely used rod-shaped fuel

elements can be used, and the fuel-element design is simple. The characteristics of the proposed reactor

show its advantages over the HTGR and supercritical-water-cooled reactors.

Different types of reactors have been considered in the course of the development of nuclear power: reactors with

organic coolant and coolant based on dissociating N2O4 as well as gas-cooled heavy-water and graphite reactors [1–4].

For different reasons, not all types of reactors were developed.

At present, gas-cooled reactors are under consideration in a variant with a Brayton cycle and graphite moderator

(HTGR). Even though a great deal of time has gone into development, active adoption of such reactors has not occurred.

Thermodynamic efficiency 45.6% is attainable in gas-cooled heavy-water reactors when the coolant is heated to 500°C and

well-developed rod-shaped fuel elements are used [5]. This is possible because fission energy, including the neutron moder-

ation energy, is fully utilized and high-pressure steam in the Rankine cycle and triple superheating of the steam are used.

A positive quality of this cycle is high dryness of the spent steam (0.93 versus 0.7 in the turbines of light-water reactors),

which lowers the production costs of turbine blades [6]. The advantages of light-water reactors can be fully realized in a

closed fuel cycle with reprocessing of spent fuel.

The aim of the present work is to show that a water-moderated gas-cooled reactor based on technical advances is

possible [5].

Construction of a Water-Moderated Gas-Cooled Reactor. A 1000-MW(t) channel thermal reactor with a moder-

ator based on a mixture of heavy and light water is considered. The lattice spacing of the channels was varied from 12 to

16 cm. The core contains 549 channels; the outer diameter of the vessel is 10 cm. The diameter of the reactor core with the

minimum channel spacing was 3.06 m, and the height of the fuel-element column in a fuel assembly was 2.2 m. Each fuel

assembly contains 59 ∅6.8-mm fuel elements. The moderator pressure was 2.5 MPa, and the pressure of the coolant (hydro-

gen) at the entry into the fuel assembly was 6 MPa. The moderator temperature was 200°C at the core entrance and 210°C at

the exit. The underheating to boiling at the core exit is 14°C.

The layout of the reactor is shown in Fig. 1. The coolant in the channel is delivered from the bottom upward by

means of integral headers [7]. A cylindrical chamber is placed above the top header; the diameter of this chamber equals the

diameter of the core and the height equals that of a fuel assembly, which makes it possible to rearrange the fuel assemblies.

The control rods are placed at the bottom of the core. An analog of this solution is the control system of a boiling-water

Atomic Energy, Vol. 116, No. 1, May, 2014 (Russian Original Vol. 116, No. 1, January, 2014)

UDC 621.039.524.46.034.3

Institute of Atomic Energy, National Nuclear Center of the Republic of Kazakhstan, Kurchatov, East-Kazakhstan Oblast. Translated fromAtomnaya Énergiya, Vol. 116, No. 1, pp. 6–10, January, 2014. Original article submitted July 18, 2013.

1063-4258/14/11601-0006 ©2014 Springer Science+Business Media New York6

V. M. Kotov, G. A. Vityuk,and A. S. Suraev

POSSIBILITIES OF GAS-COOLED

WATER-MODERATED REACTORS

ABWR [8]. The reverse arrangement, where the chamber for rearranging fuel assemblies is placed beneath the core and con-

trol system above the core, is also possible. This variant will require rooms with large heights beneath the reactor.

Energy Transfer in the Rankine Cycle. The possibilities of the technical solution proposed in the present work are

largely due to the scheme used to transfer heat from the reactor into the Rankine cycle. In a heavy-water-moderated gas-

cooled NPP, energy is transferred into the steam loop only from the coolant [3, 4]. In the process, the neutron moderation

energy and heat leaked from the channels into the moderator are completely lost. The total losses can exceed 10% of the total

energy associated with nuclear fission. In addition, the stringent requirements for minimizing the heat leakage complicate the

fuel-assembly design. In the proposed solution, the moderation energy and heat leaking from the fuel assemblies are trans-

ferred to water formed from the spent steam from the turbine (Fig. 2). The maximum steam pressure in the loop equals

20 MPa, and the maximum temperature of superheating in both heaters is 500°C.

In this scheme, all heat exchangers are implemented with counterflow of the working bodies giving up and accept-

ing energy. The scheme is supplemented with a tank that protects the fuel assemblies from the leakage of high-pressure steam

into the reactor’s gas-coolant loop. When a leak of high-pressure steam into the coolant loop appears, the steam will expand

and its pressure will decrease owing to the volume increase in the additional tank and a pressure decrease in the steam loop.

Two pumps are installed in the steam loop: the first one raises the water pressure to the moderator pressure in the reactor and

the second to the maximum pressure in the Rankine cycle. This solution simplifies the requirements for the heat-exchanger

transferring heat to the moderator.

7

Fig. 1. Layout of a water-moderated gas-cooled reactor.

Fig. 2. Scheme of energy transfer from the reactor into the Rankine cycle.

The moderator pressure and temperature must be increased in order to transfer energy from the moderator into the

Rankine cycle. For the indicated steam temperature and pressure, the energy balance in the Rankine cycle with three super-

heatings has the following form: 41.27% – water heated from 31 to 365.7°C, 14.32% – transformation of water into steam

(365.8°C), 20.11% – first superheating of steam (365.8–500°C), 12.64% – second superheating of steam (300–500°C),

11.66% – third superheating of steam (278.3–500°C). For the pressure developed by the first pump, the maximum water tem-

perature can be 275°C, and the maximum amount of thermal energy transferred by the moderator in this case will equal ~30%

of the energy released in the reactor. If the neutron moderation energy equals 7% of the total energy, then the leakage from

the fuel assembly and collectors into the moderator can reach 23% of the energy released in the fuel assembly. This can sim-

plify the design of the fuel assembly and collectors. In addition, the heat leakage from the fuel assembly decreases the

required coolant flow rate and the cost of pumping the coolant.

Rankine Cycle Characteristics. The dependence of the efficiency on the number of steam superheating loops and

the pressure in the cycle was investigated. The characteristics of the cycle were calculated using the program in [9]. Of inter-

est is the dependence of the energy fraction expended on heating water from the moment of its condensation to the steam-

formation points and steam formation on the maximum steam pressure in the cycle (Fig. 3). The increase of the energy frac-

tion expended on heating water with increasing maximum pressure is important for the technical solution under

consideration. The possibility of energy transfer into the cycle with the moderator increases with increasing pressure.

8

Fig. 3. Energy fraction expended on heating water (1) and steam formation (2)

and the sum of all fractions (3) versus the maximum pressure in the cycle.

Fig. 4. Thermodynamic efficiency (1), total efficiency (3), and steam dryness (2) at

the exit from the turbine versus the maximum steam temperature in the cycle with

maximum pressure 20 MPa.

The energy fraction going into steam formation decreases and the total energy fraction going to heating and vapor-

izing the water also decreases. The sum of the energy going into heating and steam formation as the ‘inactive’ part of the

cycle plays an important role in the conversion of the thermal energy of the cycle into mechanical energy. Accordingly,

the thermodynamic efficiency of the cycle increases with increasing steam pressure (Figs. 3 and 4).

The dependence of the efficiency and steam dryness at the turbine exit on the maximum temperature of the cycle

with pressures 20, 5.5, and 1.25 MPa at the entrance into the turbines stages is shown in Fig. 5. For different pressure ratios,

the efficiency and steam dryness change very little. Even at the minimum temperature 400°C the total efficiency is much

greater than the corresponding value for the most widely used water-moderated water-cooled reactors. Above temperature

500°C, which can be reached in the course of fuel-assembly and fuel-element optimization, the efficiency is close to that of

reactors operating in a Brayton cycle with a complicated fuel-element design and total power of the turbine and compressor

four times greater than the power of the turbine in the Rankine cycle [10].

Thermophysical Characteristics of Fuel-Assembly Variants. The thermophysical characteristics of two variants

of fuel assemblies were calculated using the program of [11]. The first one was implemented in the conventional fuel-assem-

bly scheme used in channel reactors with a metal shield and a gas-filled gap between the coolant channel and the fuel-assem-

bly casing. The second one is simpler and has no shield (Fig. 6). In this variant, the coolant flow comes into direct contact

with the fuel-assembly casing. The possibility of using fuel elements with fuel in the form of uranium dioxide and uranium

metal is examined in each fuel-assembly variant. The initial data and the computational results are presented in Table 1.

9

Fig. 5. Thermodynamic efficiency (1), total efficiency (2), steam dryness (3) and

specific volume (4) at the exit from the turbine versus the maximum pressure in

the cycle.

Fig. 6. Fuel-assembly variants for a reactor with (a) and without (b) a heat

shield: 1) moderator; 2) channel body; 3) fuel element; 4) shield.

In all variants, the coolant temperature 500°C at the fuel-assembly exit is attained without exceeding the admissible

kernel and cladding temperatures of the fuel elements. In its simplified variant, a fuel assembly transfers up to 6% of the power

into the moderator, which is less than the admissible value. In this case, the energy 1.7 MW required to pump the coolant into

the reactor is conserved.

ParameterVariant

with shield without shield

Fuel UO2 Umetal UO2 Umetal

Fuel-assembly power, kW 1821.5

Coolant Hydrogen

Coolant temperature, °C:

at fuel-assembly entry 360

at fuel-assembly exit 500

Coolant flow rate in reactor, kg/sec 473.9 362.34

Coolant pressure difference along fuel-assembly height, MPa 0.0503 0.0308

Average coolant velocity, m/sec:

at entry 98.2 75

at exit 121.8 92.3

Maximum temperature, °C:

fuel element 1239 924 1329 1021

helium gap 785 781 884 878

fuel-element cladding 684 681 789 783

Total heat flow through fuel-assembly wall, kW 3.79 3.794 108.8 109.6

Power for pumping coolant, MW ~8.57 ~6.88

10

TABLE 1. Thermohydraulic Characteristics of Channels

Fig. 7. Temperature distribution along the radius of a fuel assembly

at height 1.1 m in the cool-down regime without coolant flow.

The power that can be transferred from a fuel assembly with no thermal insulation into the moderator without pump-

ing coolant and reaching the maximum temperature of one of the fuel-assembly elements was ~10 kW, which corresponds to

the energy released by fission products 2 days after the reactor is shut down (Fig. 7).

Neutron-Physical Characteristics of the Reactor. The characteristics of a reactor differing by the content of fis-

sile material in the starting fuel and the concentration of heavy water in a mixture of heavy and ordinary water were calcu-

lated according to the program of [12]. The lattice step of the fuel assemblies in the variant with ordinary water was taken to

be 12 cm (triangular lattice). Fuel assemblies with and without a heat shield were used in the calculations and water with the

same composition as in the moderator was assumed to be in the reflector. The thickness of the reflector layer varied from 15

to 50 cm depending on the heavy-water content.

Heavy water is used because the small lattice spacing of the variant with ordinary water as the moderator makes it

more difficult to develop and maintain a system for reloading fuel assemblies. In addition, the introduction of heavy water

decreases neutron losses in protium. The mass of the requisite amount of heavy water is small and the cost of water con-

taining heavy water is much lower than that of heavy water with high enrichment. For example, the CANDU reactor with

close power capacity contains more than 200 tons of heavy water, while the reactor proposed here contains about 50 tons

of a mixture with heavy-water content about 70%. The cost of the requisite water mixture is less than 20% of the cost of

the additional yearly production of electricity because the efficiency increases from 32 to 45%.

Run Calculation. The calculation was performed using the program of [13] for a reactor operating with multiple

rearrangements of the fuel assemblies and fresh fuel replacing the parts of the fuel with maximum burnup (Fig. 8). In the vari-

ant of a fuel assembly with a heat shield and oxide fuel, the run time is 479 days and the fuel burnup is 22.06 MW·days/kg.

For the identical load of such fuel, the run time with burnup to 5.6 MW·days/kg decreases to 140 days.

The characteristics of a reactor distinguished by the moderator, fuel, lattice spacing and type of fuel assembly were

calculated. The density of fuel consisting of uranium metal was taken to be 8.8 g/cm3, which ensures that the reactor loads

are the same. Better run characteristics obtain with fuel assemblies without heat shields, water containing 95% heavy water

and lattice spacing 16 cm (Table 2). In this variant, the natural uranium utilization reaches the level characteristic for a

CANDU reactor, which is the best thermal reactor in this respect. However, CANDU burnup is approximately 4 times lower

and the amount and composition of the heavy water require much larger expenditures.

Analysis. The proposed reactor scheme and the inclusion of the reactor elements in the Rankine cycle ensures ther-

modynamic efficiency above 46% for gas heated to 500°C with maximum steam pressure 20 MPa using well-developed rod-

shaped fuel elements. The turbine in the steam loop can be constructed with minimum usage of alloyed materials in its blades,

since high steam dryness is attained at the exit. The fuel assemblies of the reactor can be implemented in a simplified scheme

with no heat shield, which lowers the production costs of this scheme and the cost of pumping the coolant and reduces the

11

Fig. 8. Variation of the content of 235U (1), 239Pu (2), and 241Pu (3) in a

reactor run with starting 235U content 2.5% in uranium dioxide fuel.

neutron losses and core size. The design of the simplified fuel assembly makes it possible to transfer into the moderator loop,

2 days after the reactor is stopped and without using the gas loop of the coolant, the energy released by the fission products

during the cool-down period.

REFERENCES

1. K. K. Polyshkin, I. Ya. Emel’yanov, P. A. Delens, et al., “Arbus nuclear power plant with organic coolant and mod-

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3. Gas-Cooled Power Reactors. Directory of Nuclear Reactors: V, VII, IAEA, Vienna (1968), pp. 243–302.

Parameter

Fuel

UO2 Umetal UO2

D2O fraction in the moderator, %

0 70 95

Fuel-assembly spacing, cm

12 16

Fuel-assembly type

with thermal insulation without thermal insulation

235U mass at run start, kg 403

Keff 1.23 1.196 1.205 1.353

Neutron leakage, % 0.6 4.2 4.2 2.55

Neutron absorption, %:

in protium 20.85 8.99 8.92 5.86

in 235U 9.18 9.91 10.09 10.6

in 238U 18.2 28.5 28.1 24.6

Nuclear fission, %:

235U 48.0 46.19 46.15 53.9

238U 2.13 2.5 2.86 1.44

Run time, days 479 462 483 972

Fuel burnup, MW·days/kg 22.06 25.37 25.74 39.6

235U remainder, kg 91 89.9 83.1 16.5

Production of, kg:

239Pu 39.3 62.3 57.8 61.3

241Pu 11.5 20.9 19.9 20.1

Uranium utilization fraction, % 0.59 0.682 0.692 1.065

12

TABLE 2. Run Characteristics of Water-Moderated Gas-Cooled Reactor Variants

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Power Systems America, Ltd (2006).

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with integral headers,” March 25, 2009, Byull. Prom. Sobstv., No. 6 (2009).

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At. Énerg., 102, No. 1, 57–63 (2007).

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12. MCNP-5.1.40 Monte-Carlo N-Particle Transport Code, LANL (2003).

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14–17 (2011).

13