Open-cycle multi-megawatt MHD space nuclear power facility

The results of calculations of the characteristics and development of a scheme and technical make-up of an open-cycle space power facility based on a high-temperature nuclear reactor for a nuclear rocket motor and a 20 MW Faraday MHD generator are presented. A heterogeneous channel-vessel IVG-1 reactor, which heated hydrogen to 3100 K, with the pressure at the exit from the reactor core up to 5 MPa, burn rate 5 kg/sec, and thermal power up to 220 MW is examined. The main parameters of the MHD generator are determined: Cs seed fraction 20%, stopping pressure at the entrance 2 MPa, electric conductivity ≈ 30 S/m, Mach number ≈ 0.7, magnetic induction 6 T, electric power 20 MW, specific energy extraction ∼4 MJ/kg. The construction of the scheme of a MHD facility with zero-moment exhaust of the working body and its main characteristics are presented. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 135–144, September, 2008.     

High-Temperature Research Reactor (30th anniversary of power startup of IVG1)

The history of the development of the first domestic high-temperature gas-cooled reactor — a unique experimental reactor for testing the core of reactors for nuclear-powered rocket motors — is briefly reviewed. The extraordinary requirements which were imposed on the reactor made it necessary to search for and implement many physical, structural, and technological innovations and to solve complicated scientific-engineering problems during development and startup. Data on the construction and characteristics of the reactor, the arrangement of the bench systems for servicing the reactor, and the results of the first startup and operation over a period of many years are presented.__________Translated from Atomnaya Energiya, Vol. 98, No. 3, pp. 163–170, March, 2005.     

Space nuclear power systems with turbomachine energy conversion

The expansion of space exploration requires the development of sufficiently powerful and reliable power facilities which can operate for a long time. Such facilities could be nuclear power systems and nuclear-powered propulsion systems with turbomachine energy conversion. The development of such systems at the present time is based on the results of work performed as part of the nuclear rocket motor program. The data presented in this article attest to the fact that our country plays a leading role in the construction of such of reactors. To maintain our leading position in space nuclear technologies, it is important to use and further develop the existing unique engineering and technological bases. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 48–50, July, 2007.     

An object-oriented approach to simulation of IRIS dynamic response

In this paper the development of an adequate modelling and simulation tool for Dynamics and Control tasks is presented. The key features of the developed simulator are: “Modularity” - the system model is built by connecting the models of its components, which are written independently of their boundary conditions; “Openness” - the code of each component model is clearly readable and close to the original equations and easily customised by the experienced user; “Efficiency” - the simulation code is fast; “Tool support” - the simulation tool is based on reliable, tested and well-documented software.To achieve these objectives, the Modelica language was used as a basis for the development of the simulator. The Modelica language is the result of recent advances in the field of object-oriented, multi-physics, dynamic system modelling. The language definition is open-source and it has already been successfully adopted in several industrial fields.The test bed for the application of the object-oriented approach has been the new generation, integral type, IRIS nuclear reactor. IRIS (International Reactor Innovative and Secure) is a pressurized light water cooled, small/medium power (335 MWe) reactor reactor, under development by an international consortium of nineteen organizations from ten countries. The preliminary design has been completed and the pre-application licensing process with the US-Nuclear Regulatory Commission (NRC) is underway.To provide the required capabilities for the analysis, specific models for the nuclear reactor components have been developed, to be applied for the dynamic simulation of the IRIS integral reactor, albeit keeping general validity for PWR plants. The following Modelica models have been written to satisfy the IRIS modelling requirements and are presented in this paper: point reactor kinetic, fuel heat transfer, control rods model, and a once-through type steam generator, thus obtaining a specific library of nuclear models and components. As far as other classical power generation plant components are concerned, the Thermo Power open library, developed at Politecnico di Milano as well, has been adopted and is briefly presented in the paper. Originally conceived for conventional, fossil-fired plants, the highly modular approach allowed to effectively reuse the models of the balance of plant systems, which have been connected to the models of the nuclear power generation process, to obtain a system simulator for the IRIS reactor.Finally, preliminary results of the code validation process and the reactor dynamics are presented.     

CERMET-SNRE堆芯物理计算分析

核火箭发动机功率高、寿命长、比冲大,在执行深空探测和星际航行任务时具有不可替代的优势。小型化是核火箭发动机的一个重要趋势,基于此提出了一种使用钨基金属陶瓷燃料的小型核火箭发动机（CERMET-SNRE）堆芯方案,并采用蒙特卡罗程序（MCNP）进行了精确建模,计算了相关物理参数。计算分析结果表明：CERMET-SNRE堆芯能谱硬,燃耗浅,后备反应性足够,功率分布合理,控制鼓与安全棒价值足够,发射掉落事故下有效增殖因数小于0.98,堆芯方案合理,满足设计要求。     

PGV-1500 steam generator. New problems and solutions

The basic questions concerning the development of a steam generator for a nuclear power plant with a VVé R-1500 reactor are presented. The basic design requirements which follow for steam generators from experience in operating analogs at nuclear power plants and taking account of the requirements for a reactor system are presented. The special features inherent to horizontal-type steam generators, which have been mastered and are used in nuclear power plants in our country, are noted. The domestic and world operating experience is taken into account in the development of the design. It is concluded that the design of the PGV-1500 steam generator satisfies the requirements for the concept of a VVéR reactor facility for a 1500 MW(e) unit of a nuclear power plant and is competitive on the world market for power-generating equipment for nuclear power plants. __________ Translated from Atomnaya énergiya, Vol. 99, No. 6, pp. 416–425, December, 2005. An erratum to this article is availabel at .     

Preliminary conceptual design of a new moderated reactor utilizing an LEU fuel for space nuclear thermal propulsion

Nuclear Thermal Rocket (NTR) propulsion is a viable and meritorious option for human exploration into deep-space because of its high thrust, improved specific impulse, well established technology, bimodal capability, and enhanced mission safety and reliability. The NTR technology has already been investigated and tested by the United States of America and Russia and the former Soviet Union. The representative Nuclear Engine for Rocket Vehicle Applications (NERVA) type reactors traditionally used Highly Enriched Uranium (HEU) fuels, shaped in hexagonal fuel element geometries because of the importance of making a high power reactor with a minimum size. Although the HEU-NTR designs are the best choice in terms of rocket performance and technical maturity, they inevitably provoke nuclear proliferation obstacles not only for all research and development activities by civilians and non-nuclear weapon states but also for potential commercialization. To overcome the security issues due to HEU, the non-proliferative, small-size NTR engine with low thrust levels of 41 kN–53 kN (9.2 klb_f ∼ 11.9 klb_f), Korea Advanced NUclear Thermal Engine Rocket utilizing a Low-Enriched Uranium fuel (KANUTER-LEU), is being designed for future generations. Its design goals are to make use of an LEU fuel for its fairly compact core, but to minimize the rocket performance sacrifice relative to the traditional HEU-NTRs. To achieve these goals, a new space propulsion reactor is conceptually designed with the key concepts of a high uranium density fuel with resistance against high heating and H₂ corrosion, a thermal neutron spectrum core, and a compact and integrated fuel element core design with protective cooling capability. In addition, a preliminary design study of neutronics and thermal-hydraulics was performed to explore the design space of the new LEU-NTR reactor concept. The result indicates that the innovative reactor concept has great potential, both to implement the use of an LEU fuel and to create comparable rocket performance, compared to the existing HEU-NTR designs.     

Concept and salient design features of a high-temperature nuclear reactor with solid coolant

The salient features of using a solid substance to cool the core of a nuclear reactor and the associated advantages and limitations are examined. Conceptual proposals concerning the core design and the arrangement of the in-reactor space of a high-temperature nuclear reactor with a solid coolant are presented. Evaluated data and some results for a model reactor are presented. __________ Translated from Atomnaya énergiya, Vol. 103, No. 3, pp. 156–161, September, 2007.     

Nuclear central heating — prospects and solutions

Large-scale development of nuclear central heating — a radical expansion of a sphere of application, large increase of cost-effectiveness and self-financing of the construction of nuclear sources of energy, increase of their fraction in the base part of the load schedule, and large-scale displacement of fossil fuel — is validated. Suggestions for a program for developing nuclear heat and power plants are examined. It is shown that the power generating units of nuclear heat and power plants must satisfy specific requirements, which requires developing specialized reactor systems. The main technical and economic characteristics of an innovative simplified boiling water reactor VK-300, specially designed for central heating power generating units, the parameters of a central heating power generating unit with VK-300, and the results of validation of investments in the construction of the VK-300 nuclear heat and power plant in Arkhangel’sk are presented. __________ Translated from Atomnaya énergiya, Vol. 103, No. 1, pp. 36–40, July, 2007.     

Heavy water reactors and nuclear power plants in the USSR and Russia: Past, present, and future

The history of the development of heavy-water nuclear reactors and the assoiated, installations in the USSR and Russia is presented. Research reactors constructed at the ITEP and under the scientific direction of the ITEP in other countires (Yugoslavia), industrial heavy-water nuclear reactors, and the Maket zero-power reactor are described. Heavy-water gas-cooled reactors for nuclear power plants are discussed in detail: the nuclear power plant with an A-1 reactor, constructed in Czechoslovakia, and the design of maximum-safety nuclear power plant. Electronuclear neutron generators and subcritical nuclear reactors and the possibility of using the for burning weapons plutonium are examined. The electronuclear neutron generator developed at the ITEP is described. State Science Center of the Russian Federation—Institute of Theoetical and Experimental Physics. Translated from Atomanaya énergiya, Vol. 86, No. 4, pp. 310–321, April, 1999.     

Bimodal solar thermionic power and propulsion system

A conceptual design of a solar power and propulsion thermionic system tied to a Rus’ carrier rocket was developed in the period 1997–1999. In the course of this work, all basic components were analyzed, a cost estimate was made, the scheme for placing the system in orbit, the operating regimes, special features of the adopted concept, as well as the mass, were determined, and the appearance of a system with a useful electrical power of at least 6 kW was formulated. The system was using, as much as possible, the accumulated knowledge on the thermionic systems developed for space nuclear power systems, the nuclear rocket motor technology, which was developed in the organizations of the Ministry of Atomic Energy and the Russian Science Center Kurchatov Institute, as well as the space-rocket technology of RKA. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 11–14, July, 2000.     

Nuclear propulsion systems

The main results of the work done at the Research and Design Institute of Electrical Technology on the development of designs and the creation of nuclear steam-producing systems (SPS) for propulsion are presented: the SPS VM-A for the first Soviet nuclear powered submarine, the SPS V-5 for the most powerful and fastest submarine with design 661 in the world team, and the designs of the first SGS MBU-40 with an integrated reactor. The basic technical solutions incorporated in the design of an integrated reactor for an advanced steam generating system are described. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 40–44, July, 2007.     

Work on domestically produced nuclear rocket motors

The history and basic results of work on nuclear rocket motors (NRMs) in Russia are presented. S.P. Korolev, I.V. Kurchatov, M. V. Keldysh, A. P. Aleksandrov et al. stood at the source of the work in this direction. Our NRM reactors used a heterogeneous scheme with a modular core. This enable component-by-component debugging of the reactor experimentally to the required degree of reliability. Ultimately, the costs of the complex tests were minimized, and technical characteristics obtained exceeded those obtained abroad. The energy-release density in the fuel elements reached 30 kW/cm³, and the temperature and rate of heating of the working body (hydrogen) reached 3100 K and 1000 K/sec, respectively. Translated from Atomnaya énerigya, Vol. 86, No. 4, pp. 296–302, April, 1999.     

Uniterm low-capacity nuclear power plant

The results of development work on an innovative low-capacity nuclear power plant (LCNPP) Uniterm together with substantiation of the need for a technical design of a reactor system and the basic design of the power plant are presented. The results of the work are used to determine the power range of the power generating unit where the main consumer qualities are guaranteed to be attained. Economic assessments of the development and operation of a two-block LCNPP under conditions in Russia are presented. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 44–48, July, 2007.     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

Adoption of Improved Algorithms for Controlling the Energy Release of a VVER-1000 Core at the Khmel’nitskii Nuclear Power Plant

Improved algorithms - a system of modern methods for controlling the energy release of the core of a VVER-1000 reactor - are described. Experience in adopting the improved algorithms in the No. 2 unit of the Khmel’nitskii nuclear power plant is presented.__________Translated from Atomnaya Energiya, Vol. 98, No. 6, pp. 414–421, June 2005.     

Solar and nuclear transport-power modules for different types of spacecraft

The results of a comparative analysis of the effectiveness of transport-power modules of different types (with solar and nuclear power systems, with electric and thermal rocket motors) for placing spacecraft from domestic cosmodromes into geostationary orbit using medium and heavy carrier rockets are presented. The main purpose of this analysis was to substantiate the regions of no-alternative application of nuclear power and power-propulsion systems for solving conventional and qualitatively new space problems in the interests of defense, science, and socioeconomic development of the country. It is concluded that at the present stage it is best not to use them together with medium-class carrier rockets. The results of a comparative analysis of the use of solar and nuclear transport modules together with heavy carrier rockets are presented for the assessment of solving the prospective problem of all-weather round-the-clock operational, highly detailed observation using an orbital system of spacecraft in a geosynchronous orbit, 3 figures. Central Scientific-Research Institute, Russian Ministry of Defense. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 15–20, July, 2000.     

Seismic response of reactor vessel internals for Korean standard nuclear power plant

This paper presents the development of seismic design criteria for the reactor vessel internals as a part of the standardization programme for the nuclear power plant in Korea. The seismic design loads of the reactor vessel internals are calculated using the reference input motions of reactor vessels taken from Yonggwang nuclear power plant units 3 and 4 which are being constructed in Korea. An overview of analysis related to the basic parameters and methodologies is presented. Also, the response of internal components to the reactor vessel motions is carefully investigated.     

The MARS nuclear reactor plant: A inherently safe, small/medium size multipurpose nuclear plant

The crucial aspects of nuclear safety: the need of a safe shutdown and of a reliable decay heat removal system have been the starting points in the development of a medium size, inherently safe, multipurpose “new” nuclear reactor: the MARS nuclear reactor.