Dni/dt = BiFi- (λi+k i)ni,
Where В i is the rate of formation of the isotope under study; F i is the relative leakage of fission products of the i-nucleotide into the primary circuit (leakage of fission products from fuel elements); λi - decay constant; r - fuel rod campaign; ki is the fraction of atoms removed from the coolant
Due to various processes, including leakage of fission products along with the coolant (due to leakage of the circuit), landing of fission products on the surface of the primary circuit, removal of fission products by the coolant purification system, burnout under the influence of neutrons.
Sources of leakage of fission products from the fuel rods of HTGR reactors are intact particles that do not have contamination, microparticles with damaged defective coatings (damage can be of technological or radiation origin), and fuel contamination of the coatings of microparticles, the matrix and the outer surface of the fuel rods.
In general, the relative leakage of fission products from fuel elements can be presented as follows:
Where ε is the proportion of particles with defective coatings; l1, l 2 - the share of fuel in the form of contamination of the coatings of microelements and the core matrix. It has been shown that the share of fuel in the coating of particles can be less than 10-6 of the total fuel content in fuel elements and, apparently, can be reduced in the future as the technology is developed; f k is the relative leakage from intact microfuel elements. Experimental work has shown that the relative leakage from intact microfuel elements coated with RuC and SiC can be taken equal to 0 for all fission products; f"k is the relative leakage from particles with damaged coating; its value significantly depends on the nature of the damage to the coating, and its determination is a difficult task; f c,f L,f"c, f"L"—relative leakage from cores and coating of fuel rods in the absence and presence of contamination in them, respectively.
Experience in developing technology for manufacturing and quality control of fuel rods shows that the relative leakage of fission products from fuel rods is mainly determined by the proportion of particles with damaged coatings and the nature of the damage. Currently, there is no data to numerically assess the effect of coating quality on leakage. Therefore, in engineering calculations we introduce
Equivalent values of the number of particles ε1",ε1" with completely damaged coatings from RuC (to characterize the leakage of gaseous fission products and halogens) and from SiC (to characterize the leakage of metallic fission products), at which the integral leakage of individual nucleides is equal to that observed in the experiment. As a rule, damage equivalence is introduced through the equilibrium relative leakage of the 133 Xe isotope. Fuel elements of HTGR reactors must meet the requirement
Leakage of inert gases and iodine. The activity of helium in the primary circuit is largely determined by the activity of gaseous fission products (g.p.d.) At the operating temperature of the fuel rods of the HTGR reactor (t< 1600 °С) относительная утечка г.п.д.,
а также I и Те может быть представлена в виде
Where E133 Xe is the equilibrium relative leakage of 133 Xe; P is a coefficient characterizing the increase in leakage Kg compared to Xe (to a first approximation, P ~ 4).
In table 4.2.1 shows the relative leakage values of the pneumatic motor. for a 1000 MW reactor, assuming that the equivalent fraction of damaged particles by 133 Xe /F133 Xe = 10-5.
The total activity of the I and Te isotopes, which characterize leakage from fuel elements, is at the level of Xe activity. However, when assessing the activity of helium with respect to iodine, it is necessary to take into account the deposition of I and Te on the surfaces of the circuit. According to various estimates, the deposition coefficient of these isotopes may lie in the range 10-3-10~4. Ignorance of data on the release, deposition and transport of fission products on the surfaces of primary circuit equipment can significantly complicate repair work.
Leakage of metal fission products Cs, Sr, Ag. When determining the leakage of Cs, Sr, it is essential to take into account the leakage of their precursors: 90 Kr, 137 Xe, as well as the influence of retention of Cs, Sr by matrix materials. It is known that Cs, Sr effectively
Absorbed by graphite used in reactors. Silver is also strongly absorbed by graphite at temperatures up to 1000-1050 °C. These isotopes, like isotopes I, are effectively deposited on the surfaces of the primary circuit. However, at present, many uncertainties remain in the behavior of these isotopes in fuel elements and in the reactor as a whole.
Table 4.2.1
Loop activity at an installation power of 1000 MW (without taking into account the influence of helium leakage and operation of the purification system)
Nucleid
Output per division, %
F∙10 5 A, Ci
Krypton 83m 114 min 0.54 0.49 22.5
85 4.48 h 1.3 0.75 82.5
85* 10.76 years 0.27 100 -
87 76.4 min 2.53 0.4 85
88 2.77 h 3.56 0.59 177
89 190 s 4.6 0.083 32
90 33 s 5.0 0.034 14.3
91 9.8 s 3.5 0.019 5.5
ΣAKr - - - ~420
Xenon 133m 2.2 days 0.16 0.65 8.7
133 5.27 days 6.69 1,550
135m 15.6 min 0.93 4.5∙10-2 3.6
135 9.16 h 6.43 0.27 145
137 3.8 min 6.18 2.2∙10-2 11
138 14.1 min 6.6 0.042 23
139 39.68 s 5.4 9.3∙10-3 4.2
140 14 s 3.5 5.5∙10-3 1.6
ΣАХе - - - ~ 750
ΣАKr+Хе ~120
* To calculate the activity of 85 Kr, it is necessary to take into account the operation of the cleaning system and the leakage of the circuit.
In table 4.2.2 shows possible values of the leakage activity of a number of radioactive fission products during normal operation of a 600 MW(e) installation, illustrating the radiation
New safety of HTGR. The assessment was carried out under the assumption that the effective fraction of destroyed particles is ~10-4, the rate of helium leakage from the circuit is ~0.2% per day; ventilation system cleaning efficiency - 10.
Table 4.2.2
Fission product leak activity under normal plant operating conditions
Fission product
Full activity, 10 7 Ci
Coolant leak
Circuit activity, Ki
Leakage from the housing, mCi/day
Leakage to atmosphere, mCi/day
I, Te 3.7 3∙10 3 Ci 3 6 0.6
90 Sr 0.24 24 Ci/year 0.5 1 0.1
Noble gases
25 - 2,4∙10 4 1,2∙10 4 2,4∙10 4
Emergency situations. Emergency situations are decisive when analyzing the radiation safety of HTGR installations. In the event of a pressure loss accident, it is most important to know the desorption coefficients of I, Sr, Cs isotopes from the surfaces of the primary circuit equipment, which accumulate (Sr, Cs) on the equipment during the entire reactor operation. In case of accidents leading to overheating of fuel elements, it is necessary to reduce the duration of overheating, since experimental studies show the resistance of microelements to short-term overheating up to a temperature of 2000-2100 °C. In addition, a large number of microelements under various operating conditions are guaranteed against simultaneous overheating and destruction of a significant portion of them.
If water enters the circuit, it can accelerate the corrosion of graphite (C+2H 2<->CH 4). However, reliable control of impurities and the operation of the treatment system suggest that the safe operation of the plants will be ensured.
Protective barriers to retain fission products in HTGR are as follows: fuel core, fuel core coatings, graphite matrix of fuel elements, primary circuit power housings, reinforced concrete containment shell.
Chapter 5. MAIN HTGR PROJECTS DEVELOPED IN RUSSIA
5.1. High-temperature gas-cooled reactor VGR-50 of the energy chemical plant ABTU-Ts-50
In 1974, a pilot industrial energy chemical plant ABTU-Ts-50 with a high-temperature gas-cooled reactor VGR-50 was developed in Russia (Fig. 5.1.1).
Rice. 5.1.1. Diagram of an energy chemical plant with a VGR-50 reactor: 1 - steam generator; 2 - gas blower; 3 - coaxial pipeline; 4 - loading capacity; 5 - VGR-50 reactor; 6 - irradiator unloading mechanism; 7 - irradiator; 8 - distribution mechanism; 9 - capacity of damaged fuel rods; 10 - separation mechanism; 11 - rejection and injection mechanism; 12 - dispensing device
This installation makes it possible to generate electricity, as well as use a nuclear reactor as a source of radiation for radiation-chemical and energy-technological processes.
For industrial installations (for example, in the chemical, metallurgical and other sectors of the economy), non-stop operation of the reactor throughout the entire process is of decisive importance.
Production cycle. This condition is met to a greater extent by a reactor with continuous fuel refueling when operating at power. VGR-50 was designed as the first pilot industrial reactor with continuous movement of fuel in the core.
During development, the following fundamental provisions were incorporated into the reactor design: the use of free backfilling of spherical fuel elements in the core; implementation of the principle of continuous passage of fuel elements into the core when the reactor is operating at power; the use of graphite as a moderator, which weakly absorbs neutrons; use of helium as a coolant; ensuring a high degree of safety due to the negative temperature coefficient of reactivity in combination with the high heat capacity of the core; compensation of reactivity with spherical absorbing elements (pels), the number of which in the core may vary during the campaign; the location of the control and protection system (CPS) rods in the side reflector and the rods directly inserted into the backfill of the ball elements; the use of existing technology for the manufacture of metal high-pressure vessels and the use for the manufacture of materials that have been mastered by industry and have proven themselves in operation in nuclear reactors; providing the ability to assemble and dismantle reactor internals.
The VGR-50 reactor (Fig. 5.1.2) consists of the following main elements: a durable high-pressure vessel with a cover on which the electromagnetic drives of the control system and loading devices are located.
Inside the housing there is a core formed by a spherical backfill of fuel rods and pellets, control rods, a basket with graphite masonry, an upper graphite reflector, an upper radiation shield, a separating device and in-reactor control (IRC) assemblies. CPS and control valve systems ensure normal operation of the reactor and obtain operational information for control, alarm and protection. The reactor has a number of independent systems for influencing reactivity: a system of control rods, a system for changing the number of fuel rods in the core during the campaign, a system for unloading fuel rods from the core.
Rice. 5.1.2. Reactor VGR-50: I - drive of the submersible compensating rod (PKS); 2 - drive of rods AR and AZ; 3 - top protection; 4 - housing cover; 5 - upper reflector; 6 - ACL rod; 7 - rod AR and AZ; 8 - active zone; 9 - basket; 10 - reactor body; II - gas collector; 12 - lower reflector; 13 - outlet pipe for fuel rods and fuel rods; 14 - housing support; 15 - dispensing device; 16 - coolant inlet-outlet pipe; 17 - channel for unloading pellets and fuel rods from the core; 18 - side reflector; 19 - in-reactor control assembly; 20 - ball inlet pipe
The VGR-50 control and protection system ensures rapid cessation of the nuclear reaction in the core, automatic maintenance of reactor power at a given level, and transfer of the reactor from one power level to another.
The reactor control and control system uses electromagnetic drives, through which the reciprocating movement of absorbing rods in the channels (24 pcs.) and insertion
Submersible compensating rods (4 pcs.) in the spherical backfill of the core.
The system for changing the number of pels in the core (SIKP) provides compensation for slow changes in reactivity during the reactor campaign. The SICP includes a container for absorbing elements, ball lines with appropriate shut-off valves and devices for detecting and removing pellets.
The VRC system allows, during operation, to measure temperature and pressure drops in individual sections of the coolant circulation path in the reactor, as well as neutron fluxes along the height and radius of the core. Thermoelectric thermometers, pneumometric tubes and direct charge detectors are used as detectors in the RRC assemblies, which make it possible to monitor the state of the reactor and the circulation circuit of the spherical elements of the installation.
The strong-dense reactor vessel VGR-50 is welded from solid forged shells and an elliptical bottom. In the upper part of the housing there are pipes for loading fuel rods and pellets into the core, in the lower part there are four pipes for supplying and removing helium coolant and four pipes on the bottom of the housing for unloading fuel rods and pellets from the reactor. The maximum outer diameter of the case is 4580 mm, the height of the case is 10,800 mm. The spherical cover is connected using pins to the reactor body. On the reactor vessel cover there are nozzles with flange connectors for installing control rod drives, loading devices and control valve assemblies.
The flange connector between the cover and the body is sealed with a torus seal made of corrosion-resistant steel with a copper gasket. The torus seal is welded to the reactor body and cover during installation. A special cavity is provided between the gasket and the torus seal, which makes it possible to control the installation seams of the torus seal for helium density and, if necessary, obtain information about the operation of the gasket during operation. The reactor body and cover are made of low-alloy chrome-molybdenum steel of the pearlitic class, satisfying the conditions of long-term heat resistance and long-term strength under neutron irradiation conditions. The selected steel is used to make
Pressurized water reactor housings and have proven themselves well under operating conditions. In the manufacture of the body and cover of the VGR-50, it was planned to almost completely use the technology for manufacturing pressurized water reactor bodies.
Inside the housing there is an upper radiation shield, an upper graphite reflector, a basket with graphite masonry and the active zone, a support shell and a distribution device. To carry out repairs and control of the internal surface of the hull, internal hull structures and assembly of the control valves, the upper protection, the upper end reflector and the basket with graphite masonry are removable.
The upper radiation shield consists of a cylindrical shell filled with absorbing material. Graphite blocks of the upper end reflector are suspended from the lower radiation protection plate using metal rods. The reactor basket is a cylindrical shell, inside of which a graphite stack consisting of side and bottom reflectors is installed on the bottom plate. The side and bottom reflectors are formed by a number of columns made of individual blocks of reactor graphite. The columns of the side reflector are connected to the shell of the basket using rods located in the peripheral holes of the graphite blocks of the side reflector. On the inner surface of the side reflector there are projections (pylons) that fit into the ball fill. The side reflector contains 24 channels for the control rods (including 12 channels in the pylons) and 12 channels for the control valve assemblies.
The lower end reflector has a central channel for unloading fuel rods and pellets from the core and a hot coolant collector, into which the coolant from the active zone and the unloading channel enters through specially made slots in the lower end reflector. The inner surface of the lower reflector is a conical surface with angles of 45 and 60°. The configuration of the internal surface of the reflector and the diameter of the discharge channel were determined based on model experiments on spherical dynamics.
A dispensing device is attached to the bottom of the reactor, which is designed to distribute fuel elements and pellets entering it from the unloading channel into four nozzles and then into the installation’s rejection and injection mechanisms. The design of the distributing device
Troystva ensures the unloading of fuel rods and pellets from the core in various emergency situations, including when a graphite block gets into the unloading channel.
The coolant circulation circuit in the reactor is designed in such a way that the temperature of the metal of the body, cover and internal structures does not exceed 350 ° C, therefore, corrosion-resistant steel, which does not become embrittled under irradiation, was chosen as the metal of the internal structures (upper protection, upper basket reflector, dispensing device) , practically does not reduce mechanical properties up to a temperature of 350 ° C and can operate in reactor conditions for 25-30 years. Considering that the maximum temperature of graphite stack blocks does not exceed 850 °C, they are made from reactor graphite, which is currently used for graphite stack blocks of RBMK reactors.
The core of the VGR-50 reactor is formed by a free backfill of fuel rods and pellets and is limited by the internal surfaces of the side and bottom reflectors and the backfill level. The fuel elements are made of microfuel elements in a graphite matrix, enclosed in a spherical graphite shell with an outer diameter of 60 mm. In the core, the accompanying movement of fuel elements and coolant is selected.
Coolant circulation in the reactor (see Fig. 5.1.1) is carried out by centrifugal gas blowers. The coolant cooled in steam generators, pumped by gas blowers, enters the reactor through four pipes located in the lower part of the housing and rises up the annular gap between the housing and the basket. The main part of this flow is directed directly into the core, a smaller part (~8%) rises up, cools the reactor cover, the upper shield and the upper reflector, and is then sent to the core. Part of the coolant entering the reactor (~10%) is directed down into the space between the bottom of the basket and the bottom of the vessel. Here, part of the lowered flow is directed into the channels of the control rods, the other part of the flow enters the unloading channel for cooling the fuel rods. Coolant flows from the core and from the discharge channel enter the hot gas collector, from which four pipes are discharged from the reactor to the steam generators and then to the gas blowers, thus closing the circulation loop.
As a result of the development and implementation of neutronics and thermal-hydraulic calculations, the following main characteristics of the reactor were obtained:
Reactor thermal power, MW 136
Coolant Helium
Pressure in the reactor, MPa 4
Coolant temperature, °C:
At the entrance to reactor 296
At the exit from reactor 810
Coolant flow through the reactor, kg/s 51
Core dimensions, m:
Diameter 2.8
Height 4.5
Specific energy intensity of the core, kW/l
Type of fuel rod and pellet Ball
Outer diameter of fuel rod and absorbing element, mm
One download campaign, eff. days 450
Average fuel burnup, MW “day/t 100,000
The developed design and the obtained calculated characteristics show the promise of using such reactors. The design of the VGR-50 reactor made it possible to test the main fundamental solutions and accumulate operating and design experience for the creation of more powerful VTGR type reactors.
5.2. Experimental industrial energy technology installation VG-400
The installation was designed for the integrated production of high-potential thermal energy (~950 °C) and electricity and can be used to carry out energy-intensive processes in a number of industries (chemical, petrochemical, etc.). The option of using high-temperature
Rature gas-cooled reactor VG-400 for a chemical-technological plant for the production of ammonia.
The installation uses an intermediate helium circuit, which prevents radioactive fission products from entering the chemical circuit and contaminating the primary circuit with chemical products (Fig. 5.2.1). The intermediate circuit increases the safety and reliability of the installation and also ensures its versatility in relation to the application of various process plants. In the VG-400 reactor, it was planned to implement the principle of a single passage of spherical fuel elements into the core (OPAZ) per campaign, which makes it possible to reduce the temperature of the fuel elements and reload them when operating at power. In 1980-1981, work was carried out on the technical design of the VG-400 reactor with the following main characteristics:
Thermal power, MW 1000
Helium temperature, °C:
At the exit from the core 950
At the entrance 350
Helium pressure, MPa 5
Core dimensions D/H, m 6.4/4.8
Fuel rod type: Ball
Fuel rod diameter, mm 60
Number of fuel rods in the core 8.10 5
Initial fuel enrichment, % 6.5
Average burn-up depth, MW.day/t 70,000
Campaign, eff. days 320
Number of cooling loops 4
Housing material
Prestressed concrete
Steam parameters:
Pressure, MPa 17.5
Temperature, °C 535
Rice. 5.2.1. Schematic diagram VG-400 installations: 1 - reactor vessel made of polyconcrete reinforced concrete; 2 - steam generator; 3 - high-temperature intermediate heat exchanger VPTO-110; 4 - loading system; 5 - technological production; 6 - steam turbine unit with generator; 7 - emergency cooling system; 8 - reactor core; 9 - unloading system; 10 - bypass valve; 11 - main circulation gas blower with shut-off valve; 12 - primary circuit safety device
Research has been carried out to clarify the parameters of the fuel loading and core design. Much attention was paid to research into emergency operating modes, substantiation of the operability of core elements, and ensuring joint operating modes of the reactor and ammonia production process unit (Fig. 5.2.2).
Calculation and experimental studies were carried out on a body made of prestressed reinforced concrete, thermal insulation, and the patterns of movement of spherical fuel rods and
The forces of introducing control rods into the ball fill were determined using 1:10 models; 1:3; 1:1 (cell), and also developed helium coolant technology, etc.
The creation of the VG-400 installation was supposed to carry out a representative test and testing of various equipment for industrial installations for energy technology purposes.
Rice. 5.2.2. Diagram of safety barriers: 1 - core with spherical fuel elements; 2 - housing made of polyconcrete reinforced concrete with a double-wall hermetic cladding; 3 - spherical fuel rod with graphite shell; 4 - microfuel element with a four-layer coating of RuC and SiC; 5 - sealed emergency shell
5.3. VGM modular reactor
Under the scientific leadership of the IAE named after. I.V. Kurchatov OKBM and VNIIAM developed modular HTGRs with spherical fuel rods for power and energy technology installations for the simultaneous production of electricity and thermal energy for energy technology purposes. The most advanced project was the VGM-200 modular reactor with a thermal power of 200 MW for an energy technology plant. The VGM project developed a number of technical solutions that meet domestic standards and conditions for the manufacture of equipment and systems.
The VGM reactor has a cylindrical active zone (without pylons) with fuel balls placed in it in the form of a free backfill, moving according to the principle of multiple circulation.
Limiting the specific power of the core to 3 MW/m 3 with a diameter of 3 m and a height of about 9 m ensures that the fuel temperature in the reactor does not exceed 1600 °C without the influence of active cooling means, including loss of coolant.
The reactor installation has one main reactor cooling circuit and one auxiliary cooling system (ACS). The coolant moves in the core from top to bottom.
The main cooling circuit includes the main circulation gas blower, high temperature heat exchanger and steam generator. A phased development of the installation was planned:
At the first stage at a temperature of 750 °C with the production of superheated steam;
At the second stage at temperatures up to 950 °C with the production of high-potential thermal energy in a high-temperature heat exchanger and superheated steam in a steam generator.
The reactor of the VGM installation has two independent systems for influencing reactivity, based on different operating principles: a rod system consisting of 24 rods (~3.4% Δk/k), and a ball reactivity compensation system (ShSKR), consisting of 22 located in the side channel reflector, with an absorber in the form of small (~6-10 mm) balls (~10.8% Δk/k).
The rod system provides compensation for rapid reactivity effects. ShSKR serves to compensate for slower
Reactivity effects such as full temperature effect and depoisoning when the plant cools down.
Loading of the absorber into the ShSKR channels is carried out under the influence of gravity from special containers. The absorber is unloaded from the channels and lifted into containers pneumatically.
In all emergency situations, including depressurization of the circuit, residual heat is removed from the core by means of a surface cooling system (SCS), consisting of three independent channels based on a passive operating principle.
Thermal energy is transferred through the reactor vessel to a water cooler located in the reactor shaft.
The design of the VGM-200 reactor is shown in Fig. 5.3.1.
Main characteristics of the VGI-200 reactor:
Thermal power, MW 200
Electric power, MW 80
Helium temperature and pressure, °C/MPa 750 (950)/7
Core dimensions D/H, m 3/9.4
Fuel cores Uranium dioxide
Fuel rod shape, outer diameter, m Ball, 60
Number of fuel elements in the core
Uranium enrichment, % 8
Heat intensity, MW.t/m 3 3.1
Estimated burnup, MW/kg 80
Campaign, days 950
The circulation rate of fuel rods through the core is 10-15
Technological production
Electricity + hydrogen
RU scheme Double-circuit
Rice. 5.3.1. Reactor VGM-200: 1 - reactor; 2 - power building; 3 - intermediate heat exchanger; 4 - steam generator; 5 - gas blower; 6 - cooling system; 7 - fuel rod circulation system; 8 - system of absorbing balls; 9 - helium purification system; 10 - relief valve; 11 - steam turbine system (energy conversion)
Chapter 6. MODULAR HIGH TEMPERATURE HELIUM REACTOR WITH GAS TURBINE GT-MGR
One of the new generation reactors that meets the requirements of the developing large-scale nuclear power industry is the modular high-temperature helium gas turbine reactor (HT-MHR). The design of this reactor is currently being developed jointly by Russian companies (Rosatom State Atomic Energy Corporation, RRC Kurchatov Institute) and the USA (ORNL, GA).
The fundamental features of GT-MGR are:
High efficiency of electricity production (efficiency ~50%);
Possibility of using high-temperature thermal energy for technological production;
Increased safety due to self-protection and the impossibility of core melting in severe accidents;
Efficient use of nuclear fuel and the ability to implement various fuel cycle options (uranium, plutonium, thorium);
Reduced thermal and radiation impact on the environment.
The introduction of GT-MGR solves many problems of nuclear energy and increases the competitiveness of nuclear power plants. The significant advantages of GT-MGR are the expansion of the use of nuclear energy in the field of industrial high-temperature technologies and the expansion of the range of countries that use nuclear energy.
In 1997, enterprises of the Russian Ministry of Atomic Energy (OKBM, VNIINM, NII NPO Luch, Siberian Chemical Combine, VNIPIET), RRC Kurchatov Institute and foreign partners developed a conceptual design of the GT-MGR. During the development of the project, another important feature of the GT-MGR was confirmed: the technical feasibility and economic efficiency of using it for the disposal of weapons-grade plutonium.
Preliminary studies carried out in Russia and the USA (GA) during 1993-1995 on the use of weapons-grade
Tonium as fuel in HTGR has shown the unique ability of this type of reactor to ensure deep (up to 90%) burning of the initially loaded plutonium during its one-time irradiation in the reactor.
Currently, a large amount of weapons-grade and energy-grade plutonium has been accumulated. Accumulated plutonium is potentially dangerous due to the possibility of unauthorized proliferation to create nuclear weapons.
Plutonium is a valuable energy product. Therefore, an effective solution to the problem of plutonium disposition is its combustion in power reactors, in particular, in GT-MHR.
The GT-MHR power plant consists of two units connected together: a modular high-temperature reactor (MHR) and a direct cycle gas turbine energy converter (GT) (Fig. 6.1). The MGR concept is based on the use of a core with a graphite moderator, fuel in the form of microsphere compacts with multilayer ceramic coatings and helium as a coolant. There are no metal structures in the core. This makes it possible to have a helium temperature at the outlet of the reactor of 850 °C or more, which ensures high efficiency of electricity production in a direct gas turbine cycle, as well as the ability to use MHR as a source of industrial high-temperature thermal energy.
The ring-type MGR core consists of 1020 hexagonal prismatic fuel blocks located in 102 columns of 10 blocks each in height.
Every year 1/3 of the fuel units are reloaded. To ensure a reactivity reserve and a negative temperature coefficient in the core, burnable absorber rods (Er 2 O 3) are used, placed in the channels of the fuel blocks. The core has a negative temperature coefficient of reactivity at any operating temperature.
Pu fuel in GT-MHR is used in the form of particles with a multilayer coating (Fig. 6.2). The plutonium oxide cores are covered with a porous buffer layer of graphite, a dense layer of pyrographite, then a layer of silicon carbide and another layer of pyrographite.
Rice. 6.1. Reactor module GT-MGR: 1 - generator; 2 - recuperator module; 3 - turbocharger; 4 - intermediate cooler module; 5 - pre-cooler module; 6 - CPS assembly; 7 - active zone; 8 - housing system; 9 - cooling system of a stopped reactor
Rice. 6.2. Components of GT-MGR fuel rods with plutonium fuel
The particles are mixed with a graphite matrix and formed into cylindrical fuel compacts in the form of a rod with a diameter of 12.5 mm and a height of 50.0 mm. They, in turn, are loaded into hexagonal prismatic graphite fuel blocks with a height of 0.8 m and a turnkey size of 0.36 m. Main characteristics of the GT-MGR reactor:
Thermal power, MW 600
Electricity production efficiency, net, % 47.2
Coolant temperature (helium), inlet/outlet, °C
Helium temperature at the compressor inlet, °C 26
Helium pressure at the reactor inlet, MPa ~7.15
Primary coolant flow, kg/s ~316
Core diameter internal/external, m 2.96/4.84
Outer diameter of radial reflector, m 7
Core height, m 8
Number of absorber rods 48
Number of Redundant Shutdown System (RSS) channels.
Burnout degree of Pu-239, %90
Average burnup, MW.day/t 650
Loaded plutonium, kg/year 250
Enrichment of loaded plutonium with Pu-239, %
Unloaded plutonium, kg/year 70
Enrichment of unloaded plutonium with Pu-239, %
Amount of weapons-grade plutonium destroyed over 60 years for one block, t
Reactor vessel material 10Х9МФБ
Housing material SPE 15Х2НМФА
Connecting body material 10Х9МФБ
Inner diameter of reactor vessels and XPE, m
Reactor vessel height, m 26
Height of the SPE body, m 37.5
Design pressure of the housing system, MPa 8
Turbomachine rotation speed, rpm 3000
Power on generator buses, MW (el) 290.45
Recuperator efficiency 0.95
Recuperator material 08Х16Н11МЗ
Heat transfer surface of the recuperator, m 2 ~66,000
GT-MGR is characterized by increased safety. The internal safety properties inherent in GT-MHR exclude melting of the core in the event of severe reactivity-type accidents and loss of coolant. Safety features and design characteristics make the GT-MGR resistant to operator errors.
GT-MGR is a new generation reactor plant. It is developed based on world-tested technologies. Most of the technical solutions for the GT-MGR installation are based on the design solutions of the Peach Bottom and Fort St. Vrain reactors (USA), worked out at the stage of their construction and operation, and on 30 years of Russian experience in designing HTGR (VG-400 reactors , VGR-50 and VGM).
Helium-cooled reactors, which were operated in the UK, USA and Germany between 1960 and 1986 (Dragon, Peach Bottom, FSV, AVR, THTR-300), demonstrated the inherent properties of this type of reactor, satisfying modern high safety requirements . Experiments conducted at the AVR reactor (Germany) showed the ability of reactors with moderate energy intensity up to 7 MW/m 3 to cool down without the intervention of active systems and operator actions. The operation of these reactors, as well as domestic radiation tests of HTGR fuel, demonstrated the ability of microfuel elements with multilayer ceramic coatings to provide deep burnup at a very high temperature of fuel element coatings, insufficient
Compatible for other types of reactor installations. When testing plutonium fuel of a similar GT-MGR composition in the Dragon and Peach Bottom reactors, the following parameters were achieved: burnup up to 750 MW.day/kg, fast neutron fluence up to 2.2. 10 21 n/cm 2, temperature up to 1400 °C (Table 6.1).
Table 6.1
Summary data on irradiation of fuel particles coated with TRISO in comparison with the requirements for GT-MGR fuel
Program
Description of fuel
Max, fast neutron fluence, × 10 25 n/m 2 (E>0.18 MeV)
Maximum burnout
Temperature range during irradiation, °C
800 MW. day/kg
TRISO-coated UO 2 fuel particles in spherical elements
Up to 20% fima
Fuel particles from UC 2 (93% enriched) with TRISO coating in compacts
Up to 78% fima
UCO fuel particles (20% enrichment) coated with TRISO in compacts
Up to 22% fima
TRISO-coated PuO 2-x fuel particles in compacts
737 MW. day/kg
Up to 1440
End of table 6.1
Germany
UO 2 fuel particles (10% enrichment) with TRISO coatings in spherical elements and compacts
Up to 14.9% fima
Fuel particles from UO 2 (4-10% enrichment) with TRISO coating in compacts
Up to 2.8
Up to 9.4% fima
"Dragon"
TRISO-coated PuO 2 fuel particles mixed with graphite in compacts
747 MW. day/kg
The GT-MGR can use various options for the nuclear fuel cycle (uranium, plutonium, thorium) and, as already mentioned, weapons-grade plutonium can be effectively burned. Efficient combustion of weapons-grade plutonium is ensured in a cycle with a single passage of fuel through the reactor without the need for reprocessing and reuse of fuel. The useful energy production per gram of plutonium loaded per unit cycle in the GT-MGR is higher than in any other reactor facility, and the composition and form of the spent fuel provide non-proliferation guarantees. Processing of weapons-grade plutonium into the form of microfuel particles can be carried out in advance, which will help improve non-proliferation guarantees.
The energy conversion system (ECS) is completely housed in the energy conversion housing. The turbomachine consists of a generator, a gas turbine, and two compressor sections vertically fixed on one shaft with magnetic bearings. The energy conversion system includes three compact heat exchangers: a highly efficient recuperator, water-cooled preliminary and intermediate coolers. The circulation diagram is shown in Fig. 6.3. Helium coolant with temperature
A temperature of 850 °C and a pressure of 7.15 MPa at the reactor outlet is supplied to a turbine located in the energy conversion housing, which drives an electric generator and high and low pressure compressors. Next, helium, through a highly efficient recuperator, having given the maximum amount of thermal energy into the cycle, enters the preliminary cooler to discharge thermal energy into the cooling tower. Relatively cold helium at 26 °C is supplied to the first section of the compressor, then to the intermediate cooler, where excess thermal energy is discharged into the cooling tower, after which it is supplied to the second section of the compressor, from where it passes through the recuperator at a pressure of 7.24 MPa and a temperature of 110 °C. Next, helium at a temperature of ~490 °C and a pressure of 7.15 MPa enters the reactor inlet.
Rice. 6.3. Schematic diagram of the installation: 1 - reactor; 2 - generator; 3 - turbine; 4, 5 - compressor; 6 - recuperator; 7 - pre-cooler; 8 - intercooler; 9 - generator heat exchanger; 10 - bypass control valve; 11 - cooling system of the mine and reactor; 12 - auxiliary heat exchanger; 13 - circulator; 14 - air heat exchanger; 15 - helium purification system; 16 - helium storage; 17 - heat exchanger; 18 - cooling tower
The scientific, design and production base developed in Russia, the USA, France and Japan for the development of gas turbomachines, highly efficient heat exchangers and heat-resistant housings provides a prepared basis for the creation of a GT-MGR installation
Options for disposal of weapons-grade plutonium in GT-MGR. Russian institutes, together with the GA company (USA), have carried out studies of the possibility of burning weapons-grade plutonium in the GT-MGR reactor. A single passage of fuel through the reactor is accepted as the main starting point. These studies included assessment of neutron and thermal-hydraulic parameters, determination of optimal overload modes, assessment of fuel stability at high burnup, and technical and economic assessments.
The joint (US and Russia) Weapons Fuel Disposition Steering Committee WGPu recommended the following initial data for a techno-economic comparison of various weapons-grade plutonium disposition options:
The amount of plutonium for disposition is 50 tons;
Disposition time - 25 years from the start of development.
The concept of plutonium disposition in GT-MHR includes the following stages:
Processing plutonium into a coated form of microfuel, eliminating its use for military purposes;
Combustion of plutonium without recycle in GT-MHR with efficient production of useful energy;
Disposal of spent fuel in geological rocks without reprocessing.
Two options for burning plutonium in GT-MGR reactors were considered.
In the first option, 50 tons of weapons-grade plutonium are converted into the form of microfuel from plutonium oxide with a multilayer coating with the production of fuel compacts and their simultaneous disposal at three-four-module stations (12 blocks in total) GT-MGR, located at three sites (four blocks each) in Seversk, Krasnoyarsk and at the Mayak chemical plant.
In the second option, 50 tons of weapons-grade plutonium are converted into microfuel from plutonium oxide with the production of fuel compacts, which are stored in intermediate warehouses and then sent for disposal to four GT-MGR units.
Long-term storage of coated particles does not cause any changes in their properties. Plutonium processed into coated particles and compacts satisfies the main criteria for the disposition of weapons-grade plutonium, including non-proliferation requirements. Fuel compacts can be stored in intermediate warehouses and supplied for combustion in the GT-MGR reactor to produce high-temperature thermal energy and electricity. The disposition diagram of weapons-grade plutonium is shown in Fig. 6.4.
Rice. 6.4. Disposition diagram of weapons-grade plutonium
The modular high-temperature helium reactor GT-MGR has a high efficiency of converting thermal energy into electrical energy. In the GT-MGR project, the amount of useful energy generated per gram of burned plutonium per single irradiation of fuel is greater than in any other reactor system (Table 6.2). Burnout of up to 90% of the original Pu-239 is achieved. The amount and isotopic composition of plutonium in spent fuel is such that this plutonium has no value for military or commercial applications.
Nya. The technology for separating plutonium from GT-MGR spent fuel has not been developed, and the low quality of the remaining plutonium does not stimulate its reprocessing.
Table 6.2
Comparison of the use of plutonium in different types of power reactors
Parameters of GT-MGR
1/3 VVER-1000 (with MOX loading in 1/3 zones)
Fast sodium reactor BN-800
Thermal power, GW
Net efficiency, % 48 33 38
Electric power, GW
0,29 0,33 0,8
Electricity generation, GW∙year, from 50 tons of weapons-grade plutonium
Share of Pu-239 destruction, % (without recycle)
The GT-MGR reactor can be effectively used to generate useful energy by burning excess weapons-grade plutonium.
The high probability of public acceptance of an alternative to HTGR for the disposal of weapons-grade plutonium is determined not only by such key points as a high level of safety, the impossibility of melting the core, and the absence of the need to evacuate the population in any conceivable emergency situations, but also high degree burning of the originally loaded plutonium - above the "spent fuel standard".
The construction of a prototype GT-MGR unit and fuel production is expected at the site of the Siberian Chemical Combine (SCC) in Seversk.
The creation of a GT-MGR installation and fuel production from weapons-grade plutonium can be one of the most effective areas of SCC conversion, since it provides jobs for specialists with their experience and knowledge, uses existing infrastructure, ensures the disposal of weapons-grade plutonium, converts plutonium production, and creates the generating plants necessary for the region power. The production of plutonium fuel and its use are carried out within the same Siberian Chemical Combine site, which eliminates the risk associated with transporting materials containing plutonium outside the protected area.
Along with solving the problem of burning weapons-grade plutonium, thanks to its excellent energy indicators and safety properties, GT-MHR on uranium (thorium) fuel can be effectively used for commercial purposes in the global energy market.
Chapter 7. EXPERIMENTAL BASIS OF HTGR
7.1. GROG Critical Stand
In the practice of domestic and foreign reactor engineering, when developing new reactors, it is envisaged to study their physical characteristics on critical builds.
One of the first domestic critical stands for the study of HTGR physics is created at the IAE named after. I.V. Kurchatov universal stand GROG. It examines issues of HTGR physics that are common to all modifications and specific, taking into account the features of various HTGR variants (primarily the VG-400 reactor). Specific feature HTGR with spherical fuel elements is a continuous reloading of nuclear fuel, and the fuel elements can pass through the reactor once or repeatedly. The main state of an HTGR operating on the OPAZ principle is equilibrium, in which fresh spherical fuel elements enter the reactor and significantly burnt-out ones exit. This leads to a large heterogeneity of the nucleic acid composition of the fuel in the fuel elements and a significant unevenness of neutron fields along the height of the reactor. It should be noted that the parameters of the core of such a reactor depend on many factors and are of a statistical nature (related to the filling of spherical fuel elements). All this leads to a large amount of research on critical assemblies.
The versatility of the upcoming research, economic and time considerations put forward the need for the stand developers to ensure the universality of the stand, i.e. the possibility of conducting experimental studies of critical systems with significantly different main parameters of the core.
In world practice, critical assemblies are used, consisting of two zones: the study zone, assembled from full-scale elements, and the ignition zone, consisting of model elements and providing the criticality condition. This concept is also used on the GROG critical test bench. At the same time, to expand the possibilities of experimental research, a wide variation in the neutronic characteristics of a system of model elements and their closeness to the properties of a system of natural elements are ensured.
Such closeness of the neutronic properties of the ignition and study zones practically eliminates boundary effects, which increases the representativeness of the study zone, since in this case the neutronics characteristics of the study zone even with a small amount full-scale elements will be identical to the properties of a large full-scale system.
Structurally, the assembly of the GROG stand is carried out as follows. A set of graphite blocks forms a cubic-shaped masonry with a face size of 450 cm. Placing various combinations of cylindrical elements in the channels of the graphite masonry provides various geometric and physical parameters of the active zone of critical assemblies. Control rods can be placed in the center channels of any graphite masonry column. At the corners of the graphite columns there are cylindrical channels with a diameter of ~2 cm, into which absorbing elements for simulating various disturbances and sensors of the measurement system can be placed. When removing graphite blocks, a fragment of the reactor under study can be inserted in their place. In Fig. 7.1.1. The characteristic composition of the critical assembly of the GROG test bench is shown, including the central studied zone of spherical fuel elements, the ignition zone and the surrounding graphite reflector.
Main parameters of the GROG stand:
Geometry of graphite masonry, cm Cube 450x450x450
Number of graphite columns 324
Core geometry Arbitrary
Maximum linear size, cm 400
Fuel Uranium
Maximum enrichment of uranium-235, % 10
Maximum number of model fuel assemblies 2304
Nuclear PC/PU ratio 200-2000
Modeled operating modes of HTGR Initial Transitional Equilibrium
Number of control rods Up to 24
Location Arbitrary
Rice. 7.1.1. GROG critical stand: 1 - spherical fuel rods of the studied area; 2 - graphite blocks of the ignition zone; 3 - graphite reflector blocks; 4 - experimental channel; 5 - control rod; 6 - oscillator; 7 - fuel assembly storage
7.2. Critical stand "Astra"
The Astra critical stand was put into operation at the Kurchatov Institute in 1980. It is intended for experimental study of the neutronic parameters of high-temperature reactors with helium coolant (HTGR). Recently, experiments have been carried out at this stand on critical assemblies simulating the physical features of reactors with an annular core, such as RBMK and GT-MGR.
Work is being carried out in the following main directions:
Determination with high accuracy of initial data on basic materials and critical assembly elements, such as graphite blocks of side and end reflectors, fuel rods, control rods, etc.;
Conducting thorough experimental studies of the neutronic characteristics of critical assemblies, simulating the features of the designed modular HTGRs, with minimal uncertainties in the measured parameters.
The experimental results are intended to verify the calculation programs used in the design of VTGR-M, for example, JAR, MCU, PNK, WIMS D4.
General form critical stand is shown in Fig. 7.2.1. The supporting structure of the critical assembly of the Astra stand is a steel body with a bottom mounted on a rigid foundation. The internal diameter of the case is 3800 mm, the wall thickness of the side surface of the case is 10 mm, the thickness of the bottom is 20 mm. For assemblies with a ring active zone, a side reflector (SR) with an outer diameter of 380 cm and a height of 640 cm and a lower end reflector (LRE), which form a cavity in the middle part, are installed at the bottom of the housing. This cavity can be filled with spherical elements, which are fuel cells(FC) when modeling GT-MGR or a mixture of fuel and absorber elements (PE) when modeling PBMR (South Africa), forming an annular core. In the central part of the assembly there is an internal reflector made of graphite blocks when modeling GT-MGR or spherical graphite elements (GE) when modeling PBMR. The space not occupied by assembly elements (cavities, channels, etc.) is filled with air under normal conditions. This section discusses the assembly simulating the GT-MGR.
In Fig. 7.2.2 shows a general view of critical assemblies with a ring active zone simulating GT-MHR, and Fig. 7.2.3 shows a cross-sectional diagram of this assembly.
Rice. 7.2.1. General view of the critical stand "Astra"
Rice. 7.2.2. General view of the critical assembly with an annular core simulating the GT-MGR reactor on the Astra stand: 1 - side graphite reflector; 2 - ring active zone; 3 - internal reflector made of graphite blocks
Rice. 7.2.3. Cross-sectional diagram of a critical assembly with an annular core, simulating the GT-MGR reactor on the Astra stand: KO1 - KO7 - compensating control elements; AZ1-AZ8 - emergency protection organs; RR - manual control body
7.3. High temperature helium loop PG-100
The PG-100 helium loop created at the Kurchatov Institute was intended for experimental research on coolant technology, fuel elements and HTGR structural materials.
A significant part of the experimental studies carried out in Russia in accordance with the program for the development of fuel rods and equipment for the primary circuit of HTGR was carried out at a complex of experimental stands, including an installation for research
Pressure test for fuel rods, helium circulation loop PG-100 at the MP reactor, ampoule channels for studying spherical fuel rods "Kashtan", ampoule channels for studying microfuel rods "Karat".
After testing, fuel rods and material samples are examined in protective chambers.
Characteristics of experimental stands
Installation for testing tightness. All fuel rods and microfuel rods intended for reactor life tests were subjected to preliminary leak-tightness tests. For this purpose, after preliminary irradiation, they were placed in a sealed heated area through which helium was blown, which carried the released gaseous fission products (GFP) to the detector. Based on the GPD activity and relative leakage (R/B = F), a conclusion is made about the suitability of the fuel rods for reactor tests. The criterion for tightness is the condition F< 10-4.
Helium loop PG-100. The loop includes the following basic technological systems and equipment elements (Fig. 7.3.1): main gas circuit with an experimental channel, gas blower unit, regenerator, heat exchangers, filter, etc., intermediate closed water circuit with water pumps, heat exchangers, etc., helium purification system with a chain of filters, a delay unit and a cryogenic unit, storage, recharge, cooling and emergency gas discharge systems.
The loop is controlled remotely and is equipped with a mimic diagram. Up to 19 spherical fuel rods or mock-ups can be loaded into the experimental channel, which are blown with a helium flow. The specified temperature of the gas and fuel elements is maintained during the experiment by regulating the flow of helium through the channel.
Rice. 7.3.1. Helium loop PG-100: 1 - reactor tank; 2 - reactor core; 3 - experimental channel; 4 - receiver; 5 - protective membrane; 6 - throttle washer; 7 - cylinder for cooling; 8 - generator; 9-11 - gas blowers; 12, 13 - heat exchangers; 14 - heater; 15 - copper oxidation unit; 16 - refrigerator; 17 - delay block; 18 - zeolite filter; 19 - regenerator; 20, 24 - metal-ceramic filter; 21 - cryogenic block; 22, 23 - heat exchangers; 25, 30 - Vacuum pump; 26 - protective membrane; 27 - receiver; 28 - throttle washer; 29 - safety valve; 31 - emergency discharge capacity
Ampoule channels “Kashtan” and “Karat”. The channels are designed for testing spherical fuel rods and microfuels in stationary modes.
The Kashtan channel contains six spherical ampoules, each of which contains one fuel rod (Fig. 7.3.2). A graphite insert is installed between the fuel rod and the ampoule, separated from them by certain helium gaps. The ampoules are installed in a special heat conductor, which is in contact with the wall of the loop channel. Heat removal from the fuel element to the cooled surface of the channel is carried out by thermal conductivity. The size of the insert and the uranium loading are selected for each fuel element based on the condition of ensuring the required temperature of the fuel element.
Rice. 7.3.2. Diagram of the “Kashtan” channel: 1 - thermocouples; 2 - fuel rod; 3 - heat pipe; 4 - gaps; 5 - liner; 6 - clamps; 7 - ampoule
The ampoules are connected by tubes to a bench system that allows for a weak flow of helium through the ampule with the removal of products released from the fuel element and determination of their composition for each fuel element using special system analysis.
Some fuel rods are equipped with thermocouples installed in the center of the fuel rod and in its cladding. This makes it possible to measure the temperature difference across the fuel element and determine its thermal conductivity.
In the “Karat” channel, microfuel elements in free filling and in pressed pellets were irradiated. The channel contains a number of isolated ampoules located along the length of the active zone. A batch of microfuel elements is placed in each of the ampoules and the specified temperature is maintained.
Main research results
Helium circulation loop PG-100. The first life tests of a batch of VGR-50 spherical graphite fuel rods began in April 1979. Basic parameters of the loop operation and testing conditions for fuel rods during two years of operation:
Reactor power, MW 27-33
Channel surrounding power, kW 1700-2800
Helium pressure, MPa 3.8-4.1
Helium consumption, g/s:
In the loop 230-250
In channel 30-100
Helium leakage from the loop (including losses when analyzing its composition), %/day
Gas temperature, °C:
In the loop 15-150
In the channel 150-600
In the working area 300-600
Fuel element temperature, °C 500-900*
Heat release in fuel rods, kW/fuel rod 0.5-1.5*
Operating time at the power of KVG-1 channels, h 13,500
Neutron fluence with E> 0.18 MeV, 10 21 n./cm 2
Burnout (fima), % 4.9-12*
GPA output (F) 10-4
*Calculated results.
The total operating time of the loop during this period is 13,500 hours. During the tests, the flow and pressure of helium were maintained fairly stable, and the total power of the reactor changed little.
The main factor influencing the energy release in the fuel rod and the temperature of the fuel rods and gas in the working section was the power release in the cells surrounding the channel. The temperature of fuel elements and gases was maintained at a given level by changing the flow of helium through the channel. This operation did not affect the operation of the gas blowers and the total gas flow through the circuit, since the channel is located on the bypass of the main circuit.
An analysis of the representativeness of the tests in relation to the operating conditions of the VGR-50 reactor being developed in the USSR shows that for most parameters, average and maximum values were achieved in the range of VGR-50 parameters (Fig. 7.3.3).
Rice. 7.3.3. Representativeness of PG-100 tests
The relative leakage of the GPD from the fuel rods, obtained by measuring the activity of helium samples from the circuit during the tests, did not exceed the permissible value for VGR-50 (F<
10~4). Это свидетельствует о работоспособности испытываемых твэлов
в достигнутом диапазоне параметров.
On the PG-100 loop, methods for purifying and monitoring the composition of helium in relation to HTGR were tested. The technological diagram of the helium purification system used on PG-100 is similar to the diagram of the unified coolant purification system VGR-50 and VG-400 (Fig. 7.3.4).
Rice. 7.3.4. Schematic diagram of technological purification of helium coolant: 1 - zeolite filter; 2 - low-temperature adsorber; 3 - mechanical filter; 4 - heater; 5 - copper oxidation unit; 6 - delay block; 7 - refrigerator; sampling routes; - gas path of the helium purification system; ===== - main circulation circuit.
Helium is purified from impurities by adsorption on zeolites with preliminary oxidation of CO and H 2 on copper oxide at a temperature of 300 °C. A cryogenic carbon block is provided to absorb nitrogen, long-lived GPD and hydrocarbon. The radioactivity of short-lived GPDs is suppressed by delay on a special carbon block located in front of the zeolites and operating at room temperature. The cleaning system is located on the bypass of the main circuit (see Fig. 7.3.4), the helium flow through it is up to 10 g/s.
Long-term operation of the purification system has shown that it provides effective purification of helium to levels characteristic of HTGR. In table 7.3.1 shows data characterizing the composition of helium during operation.
Table 7.3.1
Cleaning system status
H 2 O H 2 CO CH 4 O 2 Ar
Shutdown Up to 8
Switching on 1 1 2-10 - 110<1 1-2
The main source of impurity leakage during operation of the PG-100 is the lubrication of gas blowers.
The level of helium activity in the loop is 10-7 Ki/l (85 Kr), there is practically no radioactive contamination of the loop equipment. Near contaminated equipment (heat exchangers, delay units, mechanical filters), the dose rate is up to 10 µR/s, which ensures equipment maintenance without any restrictions.
There were no emissions of radioactivity into the room during the entire period of operation, air pollution with aerosols and gas pressure did not exceed 10~10 Ci/l. This indicates good tightness of the circuit equipment.
Ampulla channels. "Chestnut". Two Kashtan channels were tested with a set life of 1 10 4 and 1.4∙10 4 hours. The achieved burnup was 6-15% fima at a fuel rod power of 1-2 kW, a maximum fuel rod core temperature of 1000-1500 °C and fluence neutrons (2-3.6) 10 20 n./cm 2.
The results obtained during testing showed that if in the first channel of Kashtan-1 the permissible leakage of 135 Xe F = 10-4 was achieved at a burnup depth of 4-6.5%, then due to improvements in the manufacturing technology of fuel rods in the second channel such leakage is achieved with a burn-up depth of 6-8%. The effective thermal conductivity of fuel elements also increased from 15 to 20 W/(m∙K).
The Kashtan-3 channel is currently being irradiated.
"Carat". The main parameters of microfuel elements and testing conditions for microfuel in the Karat channels are as follows:
Microfuel parameter:
Core diameter, microns 400-600
T/D* 0.15/0.5
Buffer layer thickness, µm 15-60
Buffer layer density, g/cm2 1-1.4
Irradiation conditions:
Burnout depth, % 2-15
Fast neutron fluence, n./cm 2 10 20-10 21
Temperature, °C 1100-1900
Medium Helium
* T - coating thickness; D - core diameter.
About 20 experimental batches of microfuel elements with a four-layer coating on a core of UO 2 were irradiated.As a result of post-reactor studies of microfuel elements in protective chambers, the following characteristic phenomena were discovered. At a high density of the buffer layer (~1.3 g/cm 3), spear-shaped cracks appear in it, which propagate into the dense bearing layer of RuS and partially destroy it. Under certain irradiation conditions, an amoeba effect was observed in microfuel elements, in which carbon from the hot side of the RuS coating is transferred to the cold side. At temperatures above 1700 °C and a burnout depth of 12-14%, the amoeba effect leads to complete destruction of coatings.
Protective layers of dense pyrocarbon and silicon carbide, as shown by metallographic analysis, remain intact in most cases with a burn-up depth of up to 10-15% and irradiation temperatures no higher than 1300-1500 °C.
Large-scale studies of the properties of microfuels, spherical and prismatic fuel elements, including endurance tests, are carried out at RIAR.
For this purpose, research reactors SM-3, RBT-6, and Mir are used. Channels and irradiation devices, as well as stands for pre- and post-reactor fuel research, have been created.
In addition, reactor tests of spherical fuel rods and MT for the VGR-50 installation were carried out at the research reactors VVR-Ts (Obninsk) and IVV-2 (Ekaterinburg).
Comprehensive studies on the development of spherical fuel rods, microfuels and helium technology for domestic HTGRs were carried out using research reactors with loop and ampoule channels created in Russia. Lifetime values of burnup depth and neutron fluence were achieved to justify the technical design of VGR-50 while maintaining the tightness of fuel elements and the integrity of coatings in microfuel elements.
7.4. Extra-reactor gas circuit of the central gas station
Out-of-reactor loop studies of PD transport can be implemented at the experimental base in Russia (after upgrading the existing bench equipment and at new installations).
The Kurchatov Institute RRC has an off-reactor gas circuit - a circulating helium stand (CHS), created in the 1980s, which was successfully operated and is currently mothballed (2006).
The problem of radionuclide deposition can be solved using off-reactor loop installations. They are much simpler and have lower levels of radionuclide content than in-reactor ones, therefore they are cheaper to manufacture and operate.
However, out-of-reactor loops have technical limitations, the main of which is due to the fact that they use artificial PD sources, which require the creation of certain temperature conditions to maintain the required partial pressure of PD in the coolant and maintain the required chemical composition of the coolant. Therefore, in a number of off-reactor loops (KFA, Jülich), weakly irradiated samples of fuel elements or a small coolant flow from the primary circuit of the AVR reactor were used as a source of PD.
There is an ORNL concept design for an off-reactor loop for PD transport as part of the PD transport program for a steam cycle plant. The ORNL project can be considered in detail when modernizing the CGS or developing a new
Contour for the GT-MGR program. You should also take into account previous experience in the design and operation of off-reactor circuits for PD transport at KFA.
The TsGS stand was created under the HTGR development program and was used to study the dynamics of impurities for corrosion and the properties of materials depending on the temperature and duration of exposure. Other experiments were also carried out at the stand, but without radioactivity. Placing the stand in a building specialized for working with radioactive materials makes it possible to organize such work at the Central Geological Station.
In Fig. 7.4.1 shows the CGS diagram. The stand has a closed helium circuit, consisting of a gas heater, a refrigerator, a gas blower, a working chamber, two rotary elbows, a gas purification and analysis system, pipelines, fittings, etc. The pipelines of the circuit are designed for pressure up to ~5 MPa, individual components of the circuit - up to ~ 10 MPa.
During operation of the stand the following parameters were achieved:
The maximum gas pressure in the circuit is 10 MPa;
Helium consumption 10-15 g/s;
Re number ~5000;
Maximum heater power 78 kW;
The maximum temperature of helium after the heater is 1050 °C;
The helium temperature at the beginning of the working chamber is 975 °C;
The helium temperature at the end of the working chamber is 685 °C.
At the stand, an electric 4-section nichrome heater is installed directly in the helium flow. One of the four sections has automated power control to compensate for daily power fluctuations in the supply network.
The working life of the vortex gas blower installed on the stand is ~2800 hours. The gas blower housing and shaft rotation bearing are designed for long-term operation at pressures up to ~5 MPa.
The stand includes an experimental helium compressor MTs 125 with basic technical data:
Compressor type - centrifugal;
Rotation system - gas supports;
Electric motor - three-phase 220 V;
Current frequency 400 Hz;
The pressure at the compressor inlet is 50 MPa;
Mass productivity 100 g/s;
The pressure increase in the compressor is 1.53 MPa.
Installing a compressor into the bench circuit requires the development (or acquisition) of a power supply system and testing of the experimental sample itself.
Rice. 7.4.1. Scheme of the CGS stand: NTA - low-temperature adsorber; CF - zeolite filter; MOB - copper-oxide block; O - bypass; - cooling water. Measurements: T - temperature; G - helium consumption; P - pressure; ΔР - gas blower pressure.
The bench helium purification system consists of blocks: copper oxide for the oxidation of H 2 (into water) and CO (into CO 2), a zeolite filter for removing water vapor, a low-temperature (liquid nitrogen temperature) adsorber made of activated carbon for removing N 2 and CH 4.
The analysis system of the stand consists of a gas chromatograph XTM-73, a moisture meter "Baikal", a chromatograph 2210-AL-11 (France), an analyzer (H 2 O, H 2, CO 2) "Fluorit", a spectral emission analyzer "Optics".
The dynamics of impurities in a circuit without samples, the effect of heated helium with impurities on the properties of materials depending on the duration of exposure, corrosion of graphite of spherical fuel elements in a helium coolant, and adsorption of water in a helium circuit were studied at the stand.
A technique has been developed for determining helium consumption by processing the temperature and power characteristics of the heater.
Due to the single-circuit nature of the GT-MHR project currently being developed, the problems of radioactivity of the equipment being serviced and its cleaning become especially relevant. Therefore, first of all, the circular helium bench (CHS) available at the Kurchatov Institute is intended to be used to study the transport and deposition of PD in the primary circuit of the GT-MGR.
For these purposes, significant modernization of the stand is required. The stand must be equipped with a system (generator) for dosed injection of PD simulators into the helium flow. Fuel elements (compacts) that have undergone weak irradiation in the F-1 reactor of the Kurchatov Institute can also be used as a source of PD.
Removable (replaceable) working areas must be installed in the loop circuit, in which samples of the metals from which the XLPE equipment is made are supposed to be placed. Working areas must be equipped with an electric heater to create the necessary temperature conditions on the samples.
The upgraded off-reactor CGS circuit with high pressure and coolant flow will be used to conduct single tests in order to obtain data on deposition and entrainment of DP, taking into account the influence of dust, under the operating conditions of the XPS. The interaction of Cs, Ag, I and Te a with structural metals, especially with alloys, will be determined
Turbines and recuperators (where the highest levels of deposits are expected), on temperature, partial pressure, surface condition and coolant chemistry.
The adsorption/sorption/desorption data of impurities for the model description of the deposition of PD on the surface of various structures of the primary circuit of the GT-MHR are systematized as sorption isotherms. However, the currently accumulated data is clearly insufficient to confidently predict the mass transfer of radionuclides in the primary circuit of the GT-MGR, adsorption and surface diffusion on turbine elements, etc. Moreover, most of the sorption isotherms were obtained at partial pressures of the studied impurities that are orders of magnitude higher than can be expected under real conditions of normal operation of GT-MHR.
Modeling the deposition of silver from a flow onto a surface involves considering the following processes:
Evaporation and removal of silver from the surface of matrix graphite into the coolant;
Sorption and desorption of silver on the surfaces of helium circuit equipment.
Based on the listed features, the following experimental program for silver deposition is proposed:
Creation of a source for generating silver vapor;
Study of silver deposition on surfaces;
Modeling the formation of dust aerosol particles and the deposition of silver on dust particles in a circuit.
Completion of the work will allow us to obtain silver sorption isotherms for the PADLOC program.
Modeling the deposition of cesium from a flow onto the surface involves considering the following processes:
Evaporation, adsorption/desorption at the MT/coolant interface;
Adsorption/desorption of cesium at the structural material/coolant interface.
Based on the listed features, the following experimental program for cesium deposition is proposed:
Creation of a source for cesium generation;
Doctor of Technical Sciences AND I. Stolyarevsky, leading researcher at the National Research Center “Kurchatov Institute”,
Director of the CORTES Center, Moscow;
Ph.D. N.G. Kodochigov, chief designer, A.V. Vasyaev, head of department,
Doctor of Technical Sciences V.F. Golovko, chief specialist, M.E. Ganin, leading design engineer,
JSC "Afrikantov OKBM", Nizhny Novgorod
1. Introduction
The growing world demand for fuel and energy with the resource and environmental limitations of traditional energy makes it urgent to timely prepare a new energy technology that can take on a significant part of the increase in energy needs, stabilizing the consumption of fossil fuels. Russia's energy strategy for the period until 2020 defines municipal heat supply as the most socially significant and fuel-intensive sector of the economy. The demand for nuclear energy sources in the areas of electricity generation and domestic heat supply is due to the rising cost of fossil fuels and increasing energy consumption. The key factors when creating nuclear power units are the high safety of power plants and their commercial attractiveness. “The Strategy for the Development of Nuclear Energy in Russia until 2030 and for the Period until 2050,” approved by the Government of the Russian Federation, provides for heat generation from nuclear energy sources of up to 30 million Gcal/year by 2020 with an annual replacement of consumption of up to 24 billion m 3 of gas. The creation and implementation of nuclear power plants in the heat supply sector will create new generating capacities and ensure savings in natural gas for export abroad, which is a factor of geopolitical significance.
However, even the large-scale introduction of nuclear energy into the field of electric generation and municipal heat supply does not solve the problem of the growing demand for motor fuel and industrial heat. The long-term scenario for the development of nuclear energy until 2050 provides for the replacement of fossil fuels not only in the utility sector, but also in energy-intensive industries by expanding the scope of nuclear energy for the production of hydrogen, process heat, and synthetic fuel. The inevitability of mass use of new energy technologies is determined by a qualitative change in environmental requirements in the energy sector and transport.
The potential for the introduction of nuclear energy in the “non-electric” sector is determined by the volume of energy consumption of process heat by industry and is not inferior in scale to the electric power industry. In the manufacturing sector, the leaders in thermal energy consumption are the chemical industry, oil refining, and metallurgy (Table 1).
Table 1. Heat consumption by manufacturing industries (2007)
| Type of production |
Million GJ |
Million Gcal |
%
|
| Food industry |
206,4
|
49,3
|
10,8
|
| Light industry |
26,8
|
6,4
|
1,4
|
| Wood processing |
46,5
|
11,1
|
2,4
|
| Coke production |
12,1
|
2,9
|
0,6
|
| Petroleum products production |
268,8
|
64,2
|
14,1
|
| Chemical production |
492,8
|
117,7
|
25,8
|
| Production of non-metallic products |
83,7
|
20,0
|
4,4
|
| Metallurgical production |
300,2
|
71,7
|
15,7
|
| Mechanical engineering |
181,3
|
43,3
|
9,5
|
| Others |
291,8
|
69,7
|
15,3
|
| Total |
1910,4
|
456,3
|
100
|
Thus, the introduction of nuclear technologies in the heat supply of industrial processes is an urgent task that still requires its solution.
The only nuclear technology today that is truly capable of most fully solving the problem of replacing fossil fuels in industrial heat supply and transport is the technology of high-temperature modular helium reactors (HMRs).
The advantages of MGR are determined by the following factors:
The ability to heat the coolant at the exit from the core to a temperature of 1000 °C, which expands the scope of nuclear energy not only for the production of electricity and municipal heat, but also for technological purposes, including the production of hydrogen;
Possibility of using different power unit schemes: with a gas turbine cycle, with a steam turbine cycle, with a circuit for transferring high-temperature heat to technological production;
Passive principle of residual heat removal, ensuring a high level of safety, including in the event of complete loss of the primary coolant;
Ensuring a regime of non-proliferation of fissile materials, which is based on the properties of ceramic microfuels;
Low thermal impact on the environment due to the possibility of implementing efficient thermodynamic cycles for converting thermal energy into electricity (in the direct Brayton gas turbine cycle, the energy conversion efficiency can reach 50% or higher);
Possibility of combined generation of electricity and heat;
A minimum number of systems and components of a reactor plant (RP) and station when using a gas turbine cycle in the primary circuit, creating the prerequisites for reducing capital and operating costs;
Possibility of modular design of the unit with a wide range of module power (from 200 to 600 MW) and varying the AC power by a set of modules;
2. Design solutions for energy sources for industrial heat supply
Based on forecast studies of the development and needs of the energy market, pre-conceptual studies have been carried out on a prototype commercial MGR reactor plant with a unified modular helium reactor with a thermal power of ~200 MW and, on its basis, a number of energy sources for various energy technology applications.
The design basis for these developments was the world experience in creating experimental installations with a high-temperature gas-cooled reactor (HTGR), the experience of developing in Russia (more than 40 years) reactor projects with HTGR of various power levels (from 100 to 1000 MW) and purpose.
The results of the development of the GT-MGR reactor plant project with a modular helium reactor, carried out within the framework of the Russian-American program, were also used.
As part of the studies, several options for MGR for energy technology purposes were considered:
For the production of electricity and municipal heat supply, with the conversion of thermal energy of the core into electrical energy in the direct gas turbine (GT) Brayton cycle - MGR-100 GT;
For the production of electricity and hydrogen by high-temperature steam electrolysis (HES) – MGR-100 VEP;
For hydrogen production using steam methane reforming (SMR) –
MGR-100 PKM;
For high-temperature heat supply of petrochemical production (NP) – MGR-100 NP.
Each version of the MGR-100 installation consists of energy and technological parts.
The energy part is maximally unified for all options and is a power unit, including a reactor and, depending on the purpose, a gas turbine energy conversion unit (WPT) designed for electricity production, and (or) heat exchange equipment units.
The technological part of MGR-100, depending on its purpose, is either a technological installation for the production of hydrogen or high-temperature heat supply circuits that supply heat to various technological processes.
The main criteria when choosing technical solutions were to ensure high technical and economic indicators in terms of generating electricity and high-potential heat, minimizing the impact on operating personnel, the population and the environment, and eliminating radioactive contamination of the technological product.
The energy source configuration is based on the following principles.
The reactor power and its design are universal for all energy source options; only the coolant parameters differ. The choice of power level for the reactor plant (215 MW) was determined by:
The needs of the electric power industry and municipal heat supply;
The needs of industrial enterprises for high- and medium-temperature heat supply of technological processes;
Technological capabilities of domestic enterprises for the production of basic reactor plant equipment, including housings.
The reactor is modular with a core consisting of hexagonal prismatic fuel assemblies, with helium coolant, and has internal self-protection properties. Safety is ensured through the use of passive operating principles of systems. Residual heat and accumulated heat are removed from the core through the reactor vessel to the reactor shaft cooling system and then into the atmosphere using natural physical processes of thermal conductivity, radiation, and convection without exceeding the limits of safe fuel operation, including in accidents with complete loss of primary coolant , in case of failure of all active circulation means and power supply sources.
The circulation of the coolant in the loops of the primary circuit is carried out by the main circulation gas blower (MCG) or compressors of the WPT turbomachine.
The layout of all MGR-100 variants under consideration is made taking into account the requirements for safe operation of the reactor installation in all possible accidents at the nuclear power plant. Each reactor plant is located in the main NPP building, which consists of an above-ground part, which is the reactor maintenance and reloading building (central hall) and an underground containment (reactor containment shell) of low pressure, located under the central hall.
The containment houses the power equipment of the reactor plant and the equipment of the main systems important for safety. The containment is made of monolithic reinforced concrete, sealed, with internal dimensions: diameter 35 m, height no more than 35 m, capable of withstanding internal pressure of the environment up to 0.5 MPa in case of depressurization of the primary circuit of the reactor plant and/or pipelines of the secondary circuit. Containment ensures optimal use of space and volume of premises, high compactness of equipment placement, facilitation of equipment replacement and fuel reloading operations, tightness in relation to adjacent rooms of the main NPP building and the environment, heat removal to the ground in beyond design basis accidents.
The design of the primary circuit equipment has a block design. The main power equipment of MGR-100 is housed in a steel block of buildings, which consists of a vertical reactor vessel, one to three vertical buildings of WPT and heat exchange equipment, and one to three horizontal connecting buildings connecting the vertical vessels into a single high-pressure vessel (Fig. 1). The main equipment casings are similar in size to the VVER reactor vessel. Particular attention is paid to minimizing the number of external pipelines of the primary circuit.
Fig.1. Layout of reactor units: a) MGR-100 GT; b) MGR-100 VEP; c) MGR‑100 PKM; d) MGR-100 Refinery
The energy source options for MGR-100 GT and MGR-100 VEP (Fig. 2.3) provide for the use of a unified gas turbine WPT. The central place in the WPT is occupied by a turbomachine (TM), which is a vertical unit consisting of a turbocompressor (TC) and a generator, the rotors of which have different rotation speeds - 9000 rpm and 3000 rpm, respectively. Electromagnetic bearings are used as the main supports. The generator is located outside the helium circulation circuit in the air. The preliminary and intermediate WPT coolers are located around the TC. The recuperator is located in the upper part of the housing above the axis of the hot flue. Waste heat is removed from the primary circuit in the preliminary and intermediate coolers by the WPT cooling water system and further to the atmospheric air in dry fan cooling towers. It is possible to consider the option of using waste heat for heating needs and hot water supply.
Heat exchanger blocks are designed to transfer thermal energy from the reactor to the consumer in energy technology production. Depending on the working environment, the type of process and the likelihood of radioactivity entering the technological production product and contamination of equipment with radioactive products, a two- or three-circuit reactor plant design can be used.
Thus, in the plant for the production of hydrogen by the method of high-temperature electrolysis of steam (MGR-100 VEP) and the method of steam reforming of methane (MGR-100 PKM), a double-circuit scheme is used. In these processes, the main component of the process medium is water vapor. The analysis shows that in possible emergency situations with depressurization of a steam generator or high-temperature heat exchanger, the effects of the entry of hydrogen-containing products into the reactor are reliably regulated by the reactor control and protection systems.
The energy source option for supplying heat to petrochemical production (MGR-100 NP) provides for a three-circuit thermal circuit. Heat is transferred from the switchgear to the consumer through a high-temperature intermediate helium-helium heat exchanger and an intermediate helium circuit, and then to the power supply circuit. This solution limits the release of radioactivity into the network circuit, ensuring the radiation purity of the process product, as well as minimal contamination of the primary circuit with process impurities.
The main technical measures aimed at eliminating the potential danger of radioactivity entering the technological production product are the creation and maintenance of a guaranteed pressure drop (~0.5 MPa) directed towards the primary circuit, and for the MGR-100 NP variant, also the introduction of an intermediate circuit. Operational leaks of helium from the intermediate circuit to the primary circuit do not have a negative impact on the reactor plant.
2.1 Energy source MGR-100 GT for electricity production and municipal heat supply
The MGR-100 GT energy source is designed to produce electricity in a direct gas turbine cycle. The high temperature of the waste heat of the gas turbine cycle (more than 100 °C) allows it to be used for hot water supply and heat supply. In the climatic conditions of Russia, such functionality is of great importance. Evidence of this is the data on the annual consumption of natural gas for the production of electricity and heat, which amount to ~ 135 and 200 billion m 3, respectively.
MGR-100 GT can be operated in two modes: in the mode of only electricity production and in the combined mode of electricity production and municipal heat supply through waste heat recovery. Thus, in addition to a higher efficiency of electricity production, MGR-100 GT provides the potential opportunity to obtain a heat utilization factor of about 99%.
When the installation operates in combined mode, waste heat is removed to the coolant of the network circuit in network heat exchangers. In power-only mode, the grid loop is turned off and waste heat is removed to the outside air in fan-fed dry coolers.
The schematic diagram of the MGR-100 GT is shown in Fig. 2. The required temperature of the network water supplied to the consumer (150 ºС) is ensured by reducing the flow rate and increasing the pressure in the WPT cooling water circuit. In order to prevent the temperature of helium at the recuperator inlet from increasing above the permissible limits (600 °C) in the combined mode, a bypass branch with an adjustable bypass of helium of the primary circuit is organized in addition to the recuperator on the HP side (from the HPC output to the recuperator output on the HP side).
Fig.2. Schematic diagram of MGR-100 GT
The main parameters of the MGR-100 GT in the mode of electricity supply and municipal heat supply are shown in Table 2. In the combined mode, the electrical power of the installation will be 57 MW, the thermal power removed by the network water will be 154 MW.
| Table 2. Main parameters of MGR-100 GT |
|
| Parameter name |
Meaning |
| Electric power generation mode |
Combined mode |
|
215
|
215
|
|
46,1
|
25,4
|
|
558 / 850
|
490 / 795
|
| Low pressure helium temperature at the recuperator inlet, °C |
583
|
595
|
|
139,1
|
134
|
| Helium flow through the bypass from the HPC output to the recuperator output on the high pressure side, kg/s |
-
|
32,2
|
|
4,91
|
4,93
|
| Turbine expansion ratio |
2,09
|
1,77
|
| Generator/TC rotation speed, rpm |
3000/9000
|
3000/9000
|
| Cooling water flow rate WPT, kg/s |
804
|
480
|
| Temperature of network water at inlet/outlet, °C |
-
|
70 / 145
|
|
|
|
|
|
The cost of generated electricity, taking into account the beneficial use of waste heat for domestic heating purposes, is practically reduced by half, compared with the option of operating only in electricity generation mode. In this case, the economic effect of eliminating thermal emissions into the environment should be taken into account.
2.2 Energy sources MGR-100 VEP and MGR-100 PKM for hydrogen production
The transition to a hydrogen economy is based, among other things, on the creation of a technology for using HTGR energy in hydrogen production processes that have high thermodynamic and technical and economic efficiency. These processes, if possible, should exclude the consumption of fossil fuels, primarily oil and gas, which have limited reserves and are valuable raw materials for industry. Such processes include the production of hydrogen from water using the following main methods: electrolysis, thermochemical decomposition and high-temperature steam electrolysis. Their cost does not depend on constantly rising oil and gas prices, unlike, for example, the production of hydrogen from methane. At the same time, for the first stage of development of hydrogen energy, with gas prices still relatively low, processes for producing hydrogen from methane are being considered. Analysis of the requirements for the efficiency of production of consumed energy and the level of heat temperature allows us to formulate requirements for HTGR as an energy source, the main of which are:
Production of high-grade heat up to 950 °C;
No contamination of hydrogen with radioactive substances or their acceptable low level;
Low cost of hydrogen production compared to traditional methods;
High level of security of the energy technology complex.
The following are considered as the main hydrogen production processes at the stage of conceptual development of MGR-100:
High temperature electrolysis of water;
Steam reforming of natural gas (methane).
| Table 3. Main parameters of MGR-100 VEP |
|
| Parameter name |
Meaning |
| Reactor thermal power, MW |
215
|
| Useful electrical power of the generator, MW |
87,1
|
| Electricity generation efficiency (net), % |
45,7
|
| Helium temperature at the reactor inlet/outlet, °C |
553 / 850
|
| Helium consumption through the reactor, kg/s |
138
|
| Helium pressure at the reactor inlet, MPa |
4,41
|
| Turbine expansion ratio |
2,09
|
| Generator/TC rotation speed, rpm |
3000/ 9000
|
| Helium flow through the turbine, kg/s |
126
|
| Helium temperature at the WPT inlet/outlet, °C |
850 / 558
|
| SG power, MW |
22,3
|
| Helium consumption through SG, kg/s |
12,1
|
| Helium temperature at the SG inlet/outlet, °C |
850 / 494
|
| Steam capacity, kg/s |
6,46
|
| Steam pressure at the steam generator outlet, MPa |
4,82
|
Schematic diagram MGR-100 VEP for the production of electricity and superheated steam, the required parameters for the purpose of producing hydrogen by high-temperature electrolysis are presented in Fig. 3.
The basis for the MGR-100 VEP variant is a reactor plant configuration with a parallel arrangement of heat exchange loops in the primary circuit. One loop includes a reactor, a steam generating unit and a main gas generator. The other is the reactor and WPT. Thus, part of the thermal energy (~10%) generated in the reactor core is transferred to the PGB for the needs of hydrogen production, the rest is converted in the WPT into electrical energy in a direct gas turbine cycle.
Rice. 3. Schematic diagram of MGR-100 VEP
The main parameters of the installation are given in Table 3. The helium temperature at the reactor outlet is 850 °C, which does not exceed the corresponding temperature in the prototype GT-MGR reactor plant. The second circuit is designed to produce superheated steam in the steam generator (Fig. 4). Helium circulation in the PHB is carried out by the main circulation gas blower. Water supply and steam removal are made through the SG cover. Steam superheated to the required parameters is discharged through pipelines to a high-temperature electrolysis unit using solid oxide electrochemical elements, in which water vapor is decomposed into hydrogen and oxygen with the separation of these reagents. The WPP installation is supplied with electricity generated by the WPT generator.
Schematic diagram MGR‑100 PKM for generating high-potential heat for the purpose of producing hydrogen using the method of steam reforming of methane is presented in Fig. 5.
Steam reforming of methane is today the main industrially developed process and adapted for the first stage of implementation of hydrogen production technologies (together with HTGR). The current global production of hydrogen is based on it. The combination of HTGR and PCM makes it possible to reduce natural gas consumption by approximately 40%, and therefore the costs required for hydrogen production. The economic efficiency of introducing PCM is determined by the price of gas and the temperature of the heat consumed. The required heating temperature of the vapor-gas mixture must be no lower than 800 C, and a further increase in temperature has practically no effect on the efficiency of the process.
Fig.5. Schematic diagram of MGR-100 PKM
Thermal energy is removed from the reactor to the working medium of the secondary circuit (steam-gas mixture) in high-temperature heat exchangers (HHE), which are an integral part of the thermal conversion apparatus (TCA). The implementation of methane conversion (CH 4 +H 2 0 (steam) + heat → CO 2 +4H 2) occurs in the TKA according to a three-stage scheme. The steam-gas mixture (steam - 83.5%, CH 4 - 16.5%) is supplied sequentially in three stages - TKA1, TKA2 and TKA3. This determines the configuration of the heat transfer unit of the reactor plant. It consists of three separate high-temperature heat exchangers VTO 1, VTO 2, VTO 3 (Fig. 6), representing individual stages (sections) of the block. The arrangement of the WTO sections along the primary circuit coolant flow is parallel, and along the steam-gas mixture flow is sequential.
After TKA-3, the steam-gas mixture (steam-55%, CH 4, H 2, CO, CO 2 - 45%) with a high concentration of hydrogen sequentially passes through the CO 2 and H 2 O purification unit and is sent to the hydrogen separation unit. The return fraction and natural gas are mixed with superheated steam and then sent to the TKA. The circulation of helium in the primary circuit is carried out by the gas circulation system and the vapor-gas mixture by compressors.
The main parameters of the installation are given in Table 4. The helium temperature at the reactor outlet is 950 ºС.
| Table 4. Main parameters of MGR-100 PKM |
|
| Parameter name |
Meaning |
| Reactor thermal power, MW |
215
|
|
450 / 950
|
| Helium consumption through the reactor, kg/s |
81,7
|
| Helium pressure at the reactor inlet, MPa |
5,0
|
| Pressure of the vapor-gas mixture at the inlet of the heat exchangers, MPa |
5,3
|
| VTO-TKA1 |
|
| Heat exchanger power, MW |
31,8
|
|
12,1 / 43,5
|
|
350 / 650
|
| VTO-TKA2 |
| Heat exchanger power, MW |
58,5
|
| Helium/vapor-gas mixture consumption, kg/s |
22,2 / 60,9
|
| Temperature of the vapor-gas mixture at the inlet/outlet, °C |
350 / 750
|
| VTO-TKA3 |
| Heat exchanger power, MW |
125
|
| Helium/vapor-gas mixture consumption, kg/s |
47,4 / 101
|
| Temperature of the vapor-gas mixture at the inlet/outlet, °C |
350 / 870
|
|
|
|
|
Depending on the type of layout (loop or block) of the main equipment of the reactor plant, the configuration of the heat transfer block may be different. In a block layout, the main equipment of the reactor plant is connected using short pipes of the “pipe-in-pipe” type; it is advisable to also include the HCG in the heat transfer block.
2.3 Energy source of MGR‑100 refinery for petrochemical production
The MGR-100 refinery is designed to generate high-grade or medium-grade heat to meet the technological needs of petrochemical production (heating network coolants), which will save about 14% of processed oil. The design basis for it was the preliminary design of a modular reactor developed in Russia in the 1980s with a core of spherical fuel elements and a helium outlet temperature of 750 °C. The project focused on generating process heat based on the requirements of a typical oil refinery.
Fig.7. Schematic diagram of MGR-100 refinery
The schematic diagram of the MGR-100 refinery is shown in Fig. 7. Helium circulation in the first and second circuits is forced and is carried out by circulation gas blowers. The working medium of the network circuit is nitrite-nitrate salt. The main installation parameters are given in Table 5.
| Table 5. Main parameters of MGR-100 refinery |
|
| Parameter name |
Meaning |
| Reactor thermal power, MW |
215
|
| Helium temperature at the reactor inlet/outlet, °C |
300 / 750
|
| Helium consumption through the reactor, kg/s |
91,5
|
| Helium pressure at the reactor inlet, MPa |
5,0
|
| PHE power, MW |
217
|
| Helium consumption of the primary/secondary circuit through the PHE, kg/s |
91,5 / 113
|
| Helium temperature of the primary circuit at the inlet/outlet of the PHE, °C |
750 / 294
|
| Helium temperature of the secondary circuit at the inlet/outlet of the PHE, °C |
230 / 600
|
| Helium pressure of the secondary circuit at the PHE inlet, MPa |
5,50
|
|
|
|
|
The main consumers of refinery heat (~50% of the thermal power of the reactor) are tubular furnaces designed for thermocatalytic oil refining. Based on the level of heating of petroleum products in furnaces, oil refining processes are divided into three types: low-temperature (up to 400 °C), medium-temperature (up to 550 °C) and high-temperature (up to 900 °C). Heat from the MGR-100 reactor plant of the refinery is also used to cover the refinery's needs for process steam (~35% of the thermal power of the reactor) and electricity (~15% of the thermal power of the reactor).
The heat transfer unit consists of an intermediate heat exchanger (IHE), an HCH, and internal metal structures (ICH).
The PHE (Fig. 8) consists of a pipe system, a set of channels (37 pcs), a collecting chamber for “hot” helium in the intermediate circuit, elements for their fastening and sealing. The main circulation gas blower is mounted in the lower part of the PHE housing.
3 Problematic issues
As part of the completed projects, the circuit configuration and 3-D layout of the installations were developed, the parameters of the circuits and the characteristics of the main equipment were determined, a computational justification of the main components of the structure was carried out, an analysis of operational and emergency conditions, a preliminary analysis of the cost of creating and constructing the reactor plant was carried out, stages and plans for R&D were determined. Most of the required R&D, including on the reactor, turbomachine and its components, recuperator, preliminary and intermediate coolers, VKM, is currently being carried out in the scope of technological developments of the GT-MGR and MGR-T reactors.
The main issues requiring additional R&D are:
Development of manufacturability of high-temperature heat exchangers;
Justification of reactor plant safety for hydrogen production;
Development of power control algorithms for reactor plants in conjunction with process control systems;
Conducting certification tests of heat-resistant metal materials.
One of the main limitations when increasing the temperature of helium at the outlet of the reactor is the maximum permissible temperature for long-term operation of the VCM reactor. When the temperature of helium at the entrance to the core increases to 600 °C, in order to achieve an acceptable temperature of the reactor vessel material (~350 °C), it is planned to modify the design of the core in terms of heat removal to the reactor vessel cooling system.
Severe requirements are placed on gas ducts transporting a heated process medium with a temperature of up to 900 °C, which should not decrease due to heat losses, since the efficiency of the technological process depends on the temperature level.
Hydrogen production is a potential source of explosion hazard. When analyzing the safety of MGR-100, accidents in the technological part of the station or at industrial sites should be considered as initiating events. During these accidents, the release of technological raw materials or processed products is possible. From a protective action perspective, the worst safety consequences would be due to the shock wave following an explosion of these products.
One of the safety criteria should be the non-exceeding of the maximum release of explosive mixtures in technological production. The amount of emission is determined by the permissible value of excess pressure in the shock wave front, adopted for the containment shell, systems and elements of the nuclear power plant.
When analyzing such accidents, one should consider both scenarios with the possibility of an explosion in the immediate vicinity of the reactor, and ensuring safety through spatial separation of the nuclear and technological parts.
4 Conclusion
The development of MGR technology in Russia from the very beginning was aimed at using nuclear energy not only for electricity production, but also for industrial heat supply as an alternative to the use of fossil fuels.
The technology of modular HTGRs, thanks to its unique properties in terms of efficiency, safety and environmental friendliness, can provide an integrated energy supply with electricity, heat and fuel, including solving the urgent problem of cost-effective hydrogen production.
Environmentally safe and requiring low costs for creation and maintenance, low-power nuclear power plants based on HTGR can become important elements of the nuclear energy infrastructure of the current century.
The design and experimental work completed to date on modular MGR-100 variants for various energy technology applications confirm the possibility of meeting the requirements for new generation reactor plants.
Development of HTGR energy technology based on MGR-100 will significantly reduce the overall costs of the HTGR program and demonstrate the capabilities and advantages for the further commercialization of this technology.
Bibliography
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2. Energy strategy of Russia for the period until 2030. Approved by order of the Government of Russia dated November 13, 2009 No. 1715
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7. High temperature gas cooled reactor – source of energy for commercial production of hydrogen. Mitenkov F.M., Kodochigov N.G., Vasyaev A.V., Golovko V.F., Ponomarev-Stepnoy N.N., Kukharkin N.Ye., Stolyarevsky A.Ya. - Nuclear power, vol. 97, issue 6, December 2004, p. 432-446.
Russia and the United States are jointly developing a project for the nuclear power plant of the future. It will significantly surpass all previous systems in terms of safety, efficiency, and many other parameters. Nuclear energy has not yet said its last word.
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Minatom started cooperation with the Americans on this project back in 1993. And to date, a preliminary design of the reactor (and station) has been developed, and much more detailed developments have been in full swing for a long time.
And since experts see the future of nuclear energy in a new type of nuclear power plant, let’s take a closer look at how it will work.
This system is called Gas Turbine - Modular Helium Reactor (GT-MHR), and in Russian - “Gas turbine - modular helium reactor” - GT-MHR.
There are two main ideas here. A nuclear reactor cooled by helium gas and with inherent safety (that is, the higher the heating, the weaker the reaction, simply based on the “physics” of the reactor, up to a natural shutdown, without any participation of the control system) and - the shortest conversion of hot helium energy into electricity - using a gas turbine of the so-called closed Brayton cycle, with a turbogenerator and reactor placed in closed capsules underground.
No extensive pipes, pumps, turbines, or masses of other pieces of hardware above the surface. The design of nuclear power plants is greatly simplified.
Dozens of systems disappear with the wave of a magic wand. No intermediate coolants that change phase (liquid-vapor), no bulky heat exchangers, almost no paths for a possible leak of anything radioactive.
Everything is encapsulated. Moreover, even a failure of the control system does not lead to fuel melting. Everything automatically extinguishes and slowly cools down due to heat dissipation into the ground surrounding the station.
The fuel for the station is uranium oxide and carbide or plutonium oxide, made in the form of balls with a diameter of only 0.2 millimeters and coated with several layers of various heat-resistant ceramics. The balls are “poured” into the rods, which form an assembly and so on.
The physical (weight of the structure, reaction conditions) and geometric parameters of the reactor are such (relatively low energy density, for example) that in any event, even complete loss of coolant, these balls will not melt.
And the entire core is made of graphite - there are no metal structures here at all, and the heat-resistant alloy is used only in the outermost casing - the capsule.
So, even if all the plant personnel unanimously “go out to drink beer,” nothing terrible will happen to the surrounding nature - the temperature in the heart of the nuclear power plant will jump to a maximum of 1600 degrees Celsius, but the core will not melt. The reactor itself will begin to cool, releasing heat into the surrounding soil.
Diagram of the “heart” of the station. On the left is a turbine with an electric generator and heat exchangers, on the right is a reactor (illustration from gt-mhr.ga.com).
The use of helium as a coolant promises a number of advantages. It is chemically inert and does not cause corrosion of components. It does not change its state of aggregation. It does not affect the neutron multiplication factor. Finally, it is convenient to direct it to a gas turbine.
It is encapsulated together with pumps and heat exchangers and rotates exclusively on axial and radial electromagnetic bearings - rolling bearings are provided as emergency bearings.
Special mention needs to be made about heat exchangers. The helium that cools the reactor makes several “loops” in the turbine unit, giving its maximum energy to the turbogenerator. In addition, there is additional cooling of the helium with water, but in the event of any accident, the system will do without it altogether, the reactor will not melt.
The result of all these innovations is the efficiency of the plant - up to 50%, versus 32% for existing nuclear power plants, plus - much more complete production of nuclear fuel (which means less irradiated uranium and less high-level waste per megawatt-hour of energy received), simplicity of design, which means lower construction costs and easier control over the work.
And, of course, safety. The Americans write that GT-MGR is the first nuclear power plant in the world that will comply with the first safety level.
There are 4 of them in total, of which zero is the highest. 0 is fantastic. Nothing can ever happen here and in general there are no hazardous materials. The first level is the highest actually possible. With it, nuclear power plants, in theory, do not require special safety systems, since the reactor itself has an internal, structurally predetermined “immunity” from any operator errors and technical damage.
According to the Americans, the plant in Chernobyl had the third (worst) safety level, which means the system is critical to human errors or equipment malfunctions. Now many operating stations have reached safety level “2”.
OKBM writes that “The Russian Nuclear Energy Development Strategy provides for the construction of the main nuclear power plant GT-MGR and a fuel production facility for it at the Siberian Chemical Plant (Seversk, Tomsk Region) by 2010, and by 2012-2015 - the creation and commissioning the first four-module nuclear power plant GT-MGR.”
Helium circulation diagram (illustration from gt-mhr.ga.com).
The Americans, in turn, provide interesting details: since GT-MGR can consume not only uranium, but also weapons-grade plutonium, such nuclear power plants become an ideal device for its disposal, which is not only safe, but also, in a certain sense, profitable. For example, Seversk will (partially, of course) provide itself with electricity from the “reduced” Russian warheads.
And the plutonium that will be unloaded from the reactor after “work”, in terms of its parameters, is completely unpromising for hypothetical use in nuclear weapons, which is also not bad for world security.
But the United States is also interested in the project - the high thermal efficiency of the helium reactor - closed gas turbine combination is a colossal benefit, both in terms of economics and environmental protection.
It should be added that the thermal power of one such installation will be 600 megawatts, and the electrical power – 285 megawatts.
The estimated service life of the GT-MHR is 60 years. Will they have time to develop industrial fusion reactors by then, or will alternative energy become truly widespread?
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reactor" - GT-MGR. A large number of American and Russian institutes and organizations, as well as companies from France and Japan, are taking part in the creation of a fundamentally new nuclear power plant.
The operation of the station, as mentioned above, is based on a gas turbine - modular 
