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 .     

Cascade subcritical liquid-salt reactor for burning transplutonium actinides

A cascade subcritical liquid-salt reactor designed for burning long-lived components of the radioactive wastes of the nuclear fuel cycle is examined. The cascade scheme of the reactor makes it possible to decrease by a factor of three the power of the driving accelerator as compared with conventional accelerator-blanket systems of equal power. The fuel composition of the reactor consists of 20% Np, Am, Cm, and other transplutonium elements and 80% plutonium, which are dissolved in a salt melt NaF(50%)-ZrF₄(50%). For a 10 MW proton accelerator, 1 GeV proton energy (10 mA current) and subcriticality depth 0.05, the thermal power of the reactor is 800 MW, which permits burning ∼70 kg/yr Np, Am, Cm, and other transplutonium actinides, i.e., service five VVéR type reactors of equal power. __________ Translated from Atomnaya énergiya, Vol. 101, No. 2, pp. 116–125, August, 2006.     

Neutronics and pumping power analyses on the Tokamak reactor for the fusion-biomass hybrid concept

The authors aim to develop a fusion-biomass combined plant concept with a small power fusion reactor. A concern for the small power reactor is the coolant pumping power which may significantly decreases the apparent energy outcome. Thus pressure loss and corresponding pumping power were studied for a designed Tokamak reactor: GNOME. First, 3-D Monte-Carlo Neutron transport analysis for the reactor model with dual-coolant blankets was taken in order to simulate the tritium breeding ability and the distribution of nuclear heat. Considering calculated concentration of nuclear heat on the in-board blankets, pressure loss of the liquid LiPb at coolant pipes due to MHD and friction forces was analyzed as a function of fusion power. It was found that as the fusion power increases, the pressure loss and corresponding pumping power exponentially increase. Consequently, the proportion of the pumping power to the fusion power increases as the fusion power increases. In case of ～360 MW fusion power operation, pumping power required for in-board cooling pipes was estimated as ～1% of the fusion power.     

Development of a design for the GT-MHR energy conversion unit

Since 1995, the General Atomics Company (USA) and OKBM have been jointly developing a design for GT-MHR — a modular helium-cooled reactor and energy-conversion unit with a direct gas-turbine cycle. The reactor power is 600 MW, and the reactor is cooled with helium at pressure 7 MPa. The energy conversion unit consists of a gas turbomachine, a recovery unit, preliminary and intermediate heat exchangers, and a generator. The turbomachine consists of a turbine, compressor, and a generator. The shaft of the turbomachine turns at a speed of 4400 rpm. Alternative designs for the energy conversion block are analyzed, which will enable making a final choice of a variant for its configuration. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 57–63, January, 2007.     

TCB混合堆Li自冷包层MHD流动设计

从磁流体动力学ＭＨＤ压降的物理原理出发，对ＴＣＢ商用混合堆Ｌｉ自冷包层的ＭＨＤ流动方式进行了改进，提出了第一壁环向流动（平行环向磁场流动），核燃料增殖区径向流动的ＭＨＤ流动的设计，以解决混合堆为改善堆的经济性而采取提高包层核燃料富集度的途径所速来的热工，ＭＨＤ压降和安全问题。分析和数值计算结果表明，第一壁环向流动设计可以满足包层核燃料富集度从０．５％增加到１％，相应的热功率从４５００ＭＷ增加到...     

Steam generators – horizontal or vertical (which type should be used in nuclear power plants with VVER?)

The steam generator is a very important component of a nuclear power plant. Historically, vertical steam generators came to be used abroad and horizontal steam generators in our country. Both types of steam generators operate successfully in nuclear power plants and satisfactorily fulfill their functions, enabling the production of electricity. Repeated attempts to re-examine the existing concepts in one or another country have been unsuccessful because there are no convincing arguments for this. Nonetheless, the question of using a different type of steam generator is raised periodically in our country and abroad. This article briefly reviews different concepts of steam generators. Their parameters, characteristics, and thermal efficiency are compared and ways to increase the latter are analyzed. It is shown that it is impossible to choose one or the other type of steam generator without making an exhaustive study and analysis of the layout of the reactor facility and its scheme, servicing, and operation as part of a nuclear power plant. A comparative analysis of layouts of reactor facilities with different types of steam generators is made. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 127–135, September, 2008.     

Plasma MHD generator for modelling energy conversion in pulsed thermonuclear reactors

Conclsuions The construction of an experimental model for studying MHD energy conversion from a pulsed thermonuclear reactor is a realistic technical task at the present time. Doing this would permit development of a large scale MHD generator module for the typical parameters of the heated working medium in a pulsed thermonuclear reactor.In principle it is possible to obtain an efficiency of at least about 40% with a linear plasma MHD generator. The efficiency of the whole plant might be increased further by utilization of the thermal energy at the outlet of the MHD channel in traditional methods.When such an MHD generator is built difficulties with the behavior of supersonic plasma streams undergoing strong velocity reduction in a channel and the associated gasdynamic problems can clearly be solved successfully by active modification of the boundary layer and appropriate profiling of the MHD channel. Some complications may arise if a regime with time varying magnetic braking is used. Also important is the problem of the behavior of the plasma stream at large magnetic Reynolds numbers (Re_m1).The basic technological problems are these: materials for the MHD channel, cooling arrangements for the channel (especially the critical cross section of the flow path), and pumping off the boundary layer at the electrodes and preventing lithium condensation on the channel walls. Because of the small magnetic field required, construction of the magnet system will clearly not present substantial technical difficulties associated with its size.The most important physical questions as well as a number of technological questions characteristic of this problem may be investigated on a fairly simple model MHD generator with an output power level of 300–500 MW, a pulse duration of 10–20 msec, and a lithium plasma source.Translated from Atomnaya Énergiya, Vol. 39, No. 6, pp. 387–391, December, 1975.     

Particulars of the choice of radiation protection for planetary-surface nuclear power plants

The results of computational studies on choosing radiation protection for planetary-surface nuclear power plants are present. Protection on the base of a 0.4–1.5 MW(t) YaEU-100 thermionic space reactor was considered for a Martian nuclear power plant and a 0.36. and 0.6 MW(t) YaEU-25 reactor was considered for a lunar reactor. The mass/size characteristics of the radiation protection were obtained for different arrangements of the nuclear power plant on the planet — directly on the surface with protection delivered or an embankment consisting of local soil and in a shaft prepared beforehand. Translated from Atomnaya énergiya, Vol. 105, No. 2, pp. 72–79, August, 2008.     

Parametric analysis of a number of space nuclear power systems with a heat-pipe energy-conversion system

A concept for space nuclear power systems with modular construction of the thermal electric generator based on ampul-type electricity-generating elements is examined. High-temperature heat pipes (molybdenum-lithium) are used to remove heat from the reactor core; heat transmission from the core to the heat-receiving surface of the pipes is accomplished by radiation. Heat is removed by means of medium-temperature (vanadium-potassium) heat pipes. Fast reactors with uranium dicarbide as fuel and carbon-graphite materials as the structural materials are examined for a large series. The thermoelectric generator consists of maximally unified electricity-generating elements using a metallic shunt as the interelectrode insulation. The mass/size characteristics of a large series 20–200 kW are estimated on the basis of an analysis of the thermophysical and mass/size characteristics of the basic elements of space nuclear power systems. 3 figures, 3 references. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 34–39, July, 2000.     

A fusion–fission hybrid reactor with water-cooled pressure tube blanket for energy production

A fusion–fission hybrid reactor is proposed to achieve the energy gain of 3000 MW thermal power with self-sustaining tritium. The hybrid reactor is designed based on the plasma conditions and configurations of ITER, as well as the well-developed pressurized light water cooling technologies. For the sake of safety, the pressure tube bundles are employed to protect the first wall from the high pressure of coolant. The spent nuclear fuel discharged from 33GWD/tU Light Water Reactors (LWRs) and natural uranium oxide are taken as driver fuel for energy multiplication. According to thermo-mechanics calculation results, the first wall of 20 mm is safe. The radiation damage analysis indicates that the first wall has a lifetime of more than five years. Neutronics calculations show that the proposed hybrid reactor has high energy multiplication factor, tritium breeding ratio and power density; the fuel cannot reach the level of plutonium required for a nuclear weapon. Thermal-hydraulic analysis indicates that the temperatures of the fuel zone are well below the limited values and a large safety margin is provided.     

Feasibility analysis of two-phase MHD energy conversion for liquid metal cooled reactors

A two-phase MHD energy conversion unit is proposed to a liquid metal cooled fast reactor. Using supercritical CO₂ as the working fluid in the gas cycle without considering friction and heat losses, the optimized cycles efficiency is obtained, which is about 5% higher than that of the gas turbine Brayton cycle with the same regenerator/compressor configurations. Based on a simple MHD power analysis and the two-phase homogeneous flow model, the important system operational conditions were estimated. The results suggest that a liquid lead pump of at least 20% of the MHD power output is needed in order to convert the 400 MW reactor heat into electricity at the specified thermal efficiency, unless a mixture foam flow of void fraction greater than 80% is achievable at very high mixture velocity.     

Cusp-stabilized mirror-based neutron source

An optimally sized Fusion Engineering Test Facility should produce 10–20 MW of power at 2 MW/m² steady-state wall loading. Because mirror cells do not scale with size, one can choose the fusion power and wall loading free from minimum size constraints. A cusp stabilized axisymmetric mirror is seen to be ideally suited for this purpose due to excellent access, a simple coil set, and good MHD properties. We present parameters for a proof of principle experiment as well as for a neutron source facility.     

Hydrogen-Oxygen MHD Generator Applied to Pulsed Power Required in Nuclear Fusion Research

Designs of large test facilities of nuclear fusion research succeeding the current large Tokamaks such as TFTR, JET and JT-60 show that huge pulsed power is required to operate the new test facilities; 700 MW for 10 s to excite poloidal coils. The present paper proposes three steps of application of MHD power generation to fusion to provide such large pulsed power. The first step is to design and construct a small scale MHD generator which excites the Demo poloidal superconducting magnet (SCM) coil being under construction in JAERI. The operating current is 30 kA with the stored energy of 40 MJ. As the working gas of MHD generator, H2-02 combustion product is selected, seeded with 5%K. The second and third steps are to construct an intermediate MHD channel of 100 MWe and a large channel of 800 MWe. Much improved designs are obtained in the present study, compared with the previous designs. For the large 800 MW generator, the maximum magnetic field becomes 3.5 T with the load current of about 100 kA, while the stored energy in the MHD magnet is estimated to be less than 0.5 GJ which is much smaller than 5~8 GJ of planned poloidal coils. The small MHD channel designed for the Demo poloidal coil is 4 m long with the peak field of 1.8 T. The cryogenic magnet can be self-excited within 20 s. The Demo poloidal coil is charged in about 4 s.     

Comparison of the results of calculations and post-reactor investigations of VVÉR-1000 fuel elements with burnup 49 MW·days/kg

The results of calculations performed with the PINw99, TRANSURANUS (V1M1V03), and TOPRA-2 computer programs are compared with data obtained from post-reactor investigations of fuel elements which operated for four years in the No. 1 unit of the Zaparozh’e nuclear power plant with a VVéR-1000 reactor to burnup ≈ 49 MW·days/kg. The initial data are analyzed, and a comparison is made of the computed and experimental elongation of the fuel elements (49 fuel elements), the yield of gaseous fission products and the subcladding pressure (35 fuel elements), and the decrease of cladding diameter and fuel-cladding gap width. It is shown that these computer programs can be used to calculate VVéR fuel elements. __________ Translated from Atomnaya énergiya, Vol. 101, No. 6, pp. 413–420, December, 2006.     

Characteristic analysis of rotor dynamics and experiments of active magnetic bearing for HTR-10GT

A 10 MW high-temperature gas-cooled reactor (HTR-10) was constructed by the Institute of Nuclear and New Energy Technology (INET) at Tsinghua University of China. The helium turbine and generator system of 10 MW high-temperature gas-cooled reactor (HTR-10GT) is the second phase for the HTR-10 project. It is to set up a direct helium cycle to replace the current steam cycle. The active magnetic bearing (AMB) instead of ordinary mechanical bearing was chosen to support the rotor in the HTR-10GT. This rotor is vertically mounted to hold the turbine machine, compressors and the power generator together. The rotor's length is 7 m, its weight is about 1500 kg and the rotating speed is 15,000 rpm. The structure of the rotor is so complicated that dynamic analysis of the rotor becomes difficult. One of the challenging problems is to exceed natural frequencies with enough stability and safety during reactor start up, power change and shutdown. The dynamic analysis of the rotor is the base for the design of control system. It is important for the rotor to exceed critical speeds. Some kinds of softwares and methods, such as MSC.Marc, Ansys, and the transfer matrix method (TMM), are compared to fully analyze rotor dynamics characteristic in this paper. The modal analysis has been done for the HTR-10GT rotor. MSC.Marc was finally selected to analyze the vibration mode and the natural frequency of the rotor. The effects of AMB stiffness on the critical speeds of the rotor were studied. The design characteristics of the AMB control system for the HTR-10GT were studied and the related experiment to exceed natural frequencies was introduced. The experimental results demonstrate the system functions and validate the control scheme, which will be used in the HTR-10GT project.     

Evaluation of alternate power conversion unit designs for the GT-MHR

General Atomics in the USA and Experimental Design Bureau of Machine Building (OKBM) in the Russian Federation have jointly developed a nuclear power plant design, the gas turbine modular helium reactor (GT-MHR). There have been considerable improvements during the last 10 years, which resulted in a more effective, efficient and safe design. The existing design is based on a 600 MW(t) reactor cooled by helium at a pressure of about 7 MPa. The power conversion unit (PCU) uses reactor outlet helium with a temperature of 850 °C in a direct Brayton cycle to achieve the cycle efficiency of about 48%. The PCU consists of a gas turbine, a recuperator, a precooler, low-pressure and high-pressure compressors, an intercooler, and a generator. The turbomachine (TM) includes the turbine, compressors and generator mounted on a single vertical shaft. TM shaft rotation speed is 4400 rpm. The shaft of generator is connected to the turbine shaft by a flexible coupling. The required grid frequency of generated electricity is achieved by a converter. All PCU components are enclosed in a single vessel. TM uses radial and axial electromagnetic bearings (EMB) for support. Catcher bearings (CB) are provided as redundant support for the TM rotor in case of EMBs failure. Several alternative PCU designs were analyzed on the basis of current progress in technologies, new world experience, and experience accumulated in the process of GT-MHR design development. Results of these analyses will be taken into account when a final PCU design is selected.     

PWR-1000XL核蒸汽供应系统设计

国产化1000MW级压水堆核电站（PWR－1000XL）是中国核动力研究设计院拟向国内用户推荐的计划在“十五”后期开始建造的核电站方案之一。PWR－1000XL的设计寿命60年，核蒸汽供应系统的主要设计特点是：采用Performanc^ 燃料组件，换料周期18个月：堆芯平均线功率密度165．2W/cm，堆芯热工裕量大于15％，堆顶结构一体化，设置RPV顶盖事故排气系统，无测温旁路系统；稳压器容积45m^3，选用△75型蒸汽发生器和100D型主泵；采用破前漏技术，设置可燃气体控制系统；采用数字化仪表和控制系统。     

Coordination control and simulation for small nuclear power plant

Compared with nuclear electric factory, marine nuclear power plant has some particular features including smaller size, faster response, and stronger load following capacity etc. This paper focuses on marine nuclear power plant. Based on static mathematical models of some important parts such as reactor core, steam generator etc., a coordination control system is designed to implement its rapid following and response when power changes. According to the Matlab/Simulink simulation, this new scheme improves fast response capacity of the control system, which contributes to the practical system design.     

Stellarator-Mirror Based Fusion Driven Fission Reactor

The version of fusion driven system (FDS), a sub-critical fast fission assembly with a fusion plasma neutron source, theoretically investigated here is based on a stellarator with a small mirror part. In the magnetic well of the mirror part, fusion reactions occur from collision of an RF heated hot ion component (tritium), with high perpendicular energy with cold background plasma ions. The hot ions are assumed to be trapped in the magnetic mirror part. The stellarator part which connects to the mirror part provides confinement for the bulk (deuterium) plasma. Calculations based on a power balance analysis indicate the possibility to achieve a net electric power output with a compact FDS device. For representative thermal power output of a power plant (P _th ≈ P _fis = 0.5–2 GW) the computed electric Q-factor is in the range Q _el = 8–14, which indicates high efficiency of the FDS scheme.     

VBÉR-300 reactor system and power-generating units for regional nuclear power production

The basic design solutions and characteristics of the VBéR-300 reactor system for the power-generating units of 150–300 MW(e) nuclear power plants and regional nuclear heat-and-electricity plants are described. The reactor system implemented as a unit is based on the technologies and solutions used for marine nuclear power systems, which have been corroborated by experience in operating nuclear-powered icebreakers. The technical-economic advantages of floating power-generating units are substantiated. __________ Translated from Atomnaya énergiya, Vol. 102, No. 1, pp. 35–39, January, 2007.