High-temperature helium-cooled reactors are the best understood nuclear technology that can supply high-temperature heat for thermal processes for producing hydrogen. The GT-MGR reactor -an innovative international modular design of a helium-cooled reactor with a gas-turbine cycle -best meets the requirements for hydrogen production and is proposed as a basis for a nuclear energy source. In this paper, the technical aspects of the proposed application of HTGR as a source of energy for producing hydrogen are analyzed. The required parameters of the energy obtained from HTGR for the presently completed and future hydrogen-production technology are examined. The problems and additional R&D work on the use of HTGR at high helium temperatures are indicated.Today, fossil fuels are mainly used in industry, for transportation, and for generating electricity. At the current rates of consumption, the stores of fossil fuel are sufficient for no more than several hundreds of years. It would be shortsighted to use up all of the available reserves of fossil fuels just to meet the increase in demand.The most efficient method for meeting increasing energy needs could be to convert nuclear power into electricity and hydrogen as the most effective and universal energy carriers. HTGR is a nuclear technology that can supply high-temperature heat for producing hydrogen. HTGR technology is highly safe and efficiently produces electricity while keeping environmental effects to a minimum.Today, the electrolysis of water and steam conversion of methane for subsequent stages -thermochemical decomposition of water and high-temperature electrolysis of water vapor -are the main technology which has been investigated and mastered by industry for producing hydrogen using energy from HTGRs. The main requirement for the source of high-temperature heat for producing hydrogen is that the coolant be heated up to 1000°C with pressures up to 5-7 MPa.Nuclear reactor designers became interested in high-temperature helium-cooled reactors more than 40 years ago because of the new possibility for heating the helium at the reactor exit up to 1000°C and the enhanced safety of the reactor. The first successes were achieved in the mid-1960s: experimental low-power reactors were developed -Dragon in Great Britain, the Pitch-Bottom nuclear power plant in the USA, and AVR in Germany. The first two reactors operated for more than 10 yr, and the last one operated for more than 20 yr, showing reliability, high readiness and safety, low radioactive contamination of the first loop, stability in transient regimes, and the capability of heating helium up to 950°C for a long time.
Historical information concerning the development of high-temperature gas-cooled reactors in the USA and Russia is presented. The reactor facilities MHTGR (USA), , VGM (Russia), GT-MGR (Russia, USA), and at the Fort St. Vrain nuclear power plant (USA) are described. The US programs for developing innovative high-temperature nuclear reactor technologies are examined. It is shown that the Russian and US technological developments for the fuel, reactor system, energy conversion system, and fission-product transport are similar.Analysis of world energy consumption with limited resources for conventional power generation shows that intensive economic development is impossible without the establishment of large-scale nuclear power capable of supplying the energy required for a substantial part of the growth in energy needs. At the present time, the largest amount of fuel-energy resources, including the most expensive and scarce -oil and gas, is used for producing heat with diverse potentials, approximately three times more than for electricity production [1, 2]. The expansion of nuclear power into high-temperature industrial production, where the scarcity of fossil fuel is even more acutely felt than in the production of electricity, can be attained by bringing high-temperature gas-cooled reactors (HTGR) which are capable of generating high-potential heat, into the nuclear power system. HTGR Development in the USA and Russia. HTGR use fuel with specific qualities as compared with other types of reactors; it contains spherical fuel pellets or fuel blocks containing fuel compacts. The foundation of the fuel composition is microfuel containing a spherical kernel with ceramic multilayer coatings, which are the main barriers for confining fission products. The first US facility with an experimental reactor, developed by the General Atomics Company, was introduced in 1967 at the Peach Bottom 115 MW(t) nuclear power plant. The objective was to demonstrate the possibilities of using HTGR for commercial production of electricity with high-temperature steam and to obtain the experience required for developing a more powerful nuclear power plant. The problem of developing a reactor with high fuel burnup (75 MW·days/kg) was posed at the same time. Thirty-three fuel modifications, including the fuel rods at the Fort St. Vrain nuclear power plant, were tested [3].The pilot nuclear power plant at Fort St. Vrain with a high-temperature gas-cooled reactor was put into operation in 1976. The main equipment in the first loop of the facility consisted of a 842 MW(t) reactor with a core comprised of prismatic fuel assemblies, steam-generators, and axial gas blowers with steam-turbine drive placed beneath the core in a vessel
Currently, nuclear society has recognized that further development of Russian nuclear power industry will be carried out under the two-component framework represented with thermal and fast reactors operating within the total closed fuel cycle. Several scenarios of the nuclear fuel cycle closure within Russian two-component nuclear power system (NPS) have been currently discussed among the industry specialists. These scenarios are based on various philosophies of the NFC closure (centralized NFC closed by thermal and fast reactors fuel or station NFC for closure of the fuel cycle of BREST reactors only), various fast reactor technologies and various types of the fuel used: UOX/MOX or mixed nitride U-Pu. The key difference is in different understanding of fast reactors role in NPS fuel cycle closure. This results in varying requirements to fast reactors and the terms of implementation of such requirements. This report focuses on the role of fast sodium reactors as a system-forming component of the future two-component nuclear power system operating within the total fuel cycle with thermal reactors.
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