Economic efficiency of power generation units based on a modular helium reactor

Economic aspects and the feasibility of introducing modular helium reactors (MHR) for power generation and for production of hydrogen under competitive market conditions are examined. A dynamic balance model is used, which makes it possible to estimate the equilibrium (market) costs of electrical energy and hydrogen, rates of growth of production, and the characteristics of the resulting profit. It is shown that using a gas-turbine modular helium reactor for production of electrical energy is clearly economically efficient for the current price of natural gas. The efficient production of hydrogen in a MHR-fuel system is possible only for significantly higher natural gas prices, which have already been reached in the world market. Translated from Atomnaya énergiya, Vol. 105, No. 5, pp. 291–296, November, 2008.     

高温气冷堆氦气轮机基本特性研究

高温气冷堆氦气轮机循环被认为是将来核能发电领域中最有潜力的方案之一。首先对高温堆氦气轮机循环进行分析和优化 ,然后着重从热力学和气体动力学角度研究氦气轮机的基本特性。结果表明 ,氦气轮机有两个主要设计特点不同于通常的燃气轮机 :一个是叶片级数多 ;另一个是叶片高度低 ,这些特性分别由氦气的物性和闭式循环的高压所导致。     

An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A&M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR.This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900–1000 °C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900–1000 °C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and discussed.     

The modular high temperature gas-cooled reactor (MHTGR) in the U.S.

The MHTGR is an advanced nuclear reactor concept being developed in the USA, under a cooperative program involving the U.S. Government, the nuclear industry, and the utilities. As its objective, this program is developing a safe, reliable, and economic nuclear power option for the USA, and the other nations of the world to consider in meeting their individual nationalistic electrical generation or process heat needs by the turn of the century. The design is based on a concept of modularization that can meet the various power needs by combining any number of 350 MW(t) reactor modules in parallel with a selected number of turbine plants in a variety of arrangements. Basic HTGR features of ceramic fuel, helium coolant, and graphite are sized and configured to provide a low power density core with passive safety features such that no operator action or external source of power is needed for the plant to meet 10CFR100 or Protective Action Guidelines limits at the 425 m site boundary. This precludes the necessity to plan for the evacuation or sheltering of the public during any licensing basis event. The safe behavior of the reactor plant is not dependent upon operator action and it is insensitive to operator error. The Conceptual Design is presently being vigorously reviewed by the U.S. Nuclear Regulatory Commission (NRC). A safety evaluation report and a licensability statement are scheduled for issuance by the NRC in January 1988.     

Multi-physics steady-state analysis of OECD/NEA modular high temperature gas-cooled reactor MHTGR-350

This paper presents the application results of MCS/GAMMA+ to multi-physics analysis of OECD/NEA modular high temperature gas-cooled reactor (MHTGR) benchmark Phase I Exercise 3. It is a part of international R&D efforts lead by the Next Generation Nuclear Plant (NGNP) US project to improve the neutron-physics and thermal-fluid simulation of (high temperature gas-cooled reactors) HTGRs, one of the next generations of safer nuclear reactors. Accurate and validated analysis tools are indeed a crucial requirement for safety analysis and licensing of nuclear reactors. To guide this effort, a numerical benchmark on the MHTGR was created by the NGNP project and formally approved in 2012 for international participation by the OECD/NEA. The benchmark defines a common set of exercises and the comparison of solutions obtained with different analysis tools is expected to improve the understanding of simulation methods for HTGRs. The coupled neutronics/thermal-fluid solution presented in this paper was obtained with the neutron transport Monte Carlo code MCS developed by Ulsan National Institute of Science and Technology and the thermal-fluid code GAMMA+ developed by Korean Atomic Energy Research Institute. The purpose of this paper is to present the GAMMA+/MCS coupled system, the calculation methodology, and the obtained solutions.     

Design methodology for insulating and cooling of a small modular reactor head by high temperature structural-thermal-fluid analysis

The design of a sodium-cooled fast reactor (SFR) head can be complicated due to its shape and functions. The head is a component placed in the pressure boundary to shield nuclear radioactive radiation. At the same time, it needs to seal the reactor vessel, support penetrating components, and minimize heat losses. This paper presents a new insulating and cooling design concept of a small SFR head. For a new design, this study shows a comprehensive design approach considering fluid-thermal-structural computations. The interactive design approach refers to dependent simulation steps of three-dimensional (3D) thermal-structural, one-dimensional (1D) heat-transfer, and 3D computational fluid dynamics (CFD) analysis. This multi-domain approach was applied to the head of the large sodium integral effect test facility called sodium test loop for safety simulation and assessment (STELLA-2). And the STELLA-2 head design was proposed as a thick plate with a sandwich type of insulation, cooling the perimeter annulus of the round head-top surface. For the structural design, the ASME design code was utilized, and the head temperature of 346?°C was calculated as its initial design temperature target. In an axial heat-transfer mode from the in-vessel to the head, a 1D finite element model gave 57 and 75 mm insulation thicknesses with a thermal conductivity of 0.07 W/m/K. The cooling effectiveness of the proposed head design was shown through a commercial CFD package.     

Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants.Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.     

间接氦气透平循环高温堆紧急停堆动态仿真

通过对10 MW高温气冷堆氦气透平发电装置(HTR-10GT)的堆芯、热交换器和透平压气机组等主要设备的数学建模和程序编制,初步建立起了一套模拟该装置瞬态特性的仿真程序.通过对该装置于5s时刻堆内引入0.1$阶跃正反应性引发的紧急停堆事故的瞬态模拟,初步验证了该装置紧急停堆预案设置的安全性和合理性,证明了旁路快开阀的设...     

The high temperature gas-cooled reactor as a source of high temperature process heat

The nuclear reactor has established itself as a future major supplier of electrical energy. The industrial market for other forms of energy, however, is almost as large and represents a new potential for the use of nuclear reactors. The high temperature gas-cooled reactor (HTGR) has been developed for commercial application in the electric power generation field. Since the HTGR is capable of delivering process heat in the temperature range of 1000–1500°F, it has many applications that would not be possible at the lower operating temperatures of water-cooled reactors. This paper briefly summarizes the development of the HTGR and outlines its salient technical features. Modifications to the reactor that enable it to be used as a process heat source are discussed. Specific applications are developed for coal gasification, steelmaking, and hydrogen production.     

Conceptual design of a helium heater for high temperature applications

The Karlsruhe Advanced Technologies Helium Loop (KATHELO) has been designed for testing divertor modules as well as qualifying materials for high heat flux, high temperature (up to 800 °C) and high pressure (10 MPa) applications. The test section inlet temperature level is controlled using a process electrical heater. To cope with the extreme operating conditions, a special design of this unit has been proposed. In this paper the conceptual design of the unit will be presented and the impact of the coupling between the cold and hot helium gas on the overall efficiency of the loop will be investigated. The detailed thermal-hydraulic analysis of the feed through of the hot helium into the low temperature pressure vessel using ANSYS CFX will be presented. The impact of the design choices on the overall energy budget of the loop will be analyzed using RELAP5-3D.     

Burnable poison for reactivity management in a very high temperature reactor

A physics study has been performed to find the optimal application of burnable poison (BP) for an excess reactivity management in a 600 MWth block-type very high temperature reactor (VHTR). Six potential BP materials (B, Gd, Er, Eu, Sm, Dy) were evaluated for an equilibrium cycle in terms of the major core performance parameters such as the burnup reactivity swing, discharge burnup, fuel and moderator temperature coefficients, and fuel temperature. In addition to the conventional 6-hole BP loading in a fuel block, a 12-hole BP application was also considered in order to accelerate the depletion rate of the BPs. The self-shielding effect of the BP was optimized for a successful management of the excess reactivity by changing the effective diameter of the BP region. Additionally, a mixed BP application of Gd and Er was proposed to make better use of the two BP materials in terms of the reactivity swing and the core temperature coefficients. It has been demonstrated that the burnup reactivity swing over a long cycle period can be reduced from ∼15,000 pcm to 3000–5000 pcm with only a small reduction of the discharge fuel burnup.     

Thermohydraulics in a high temperature gas-cooled reactor prestressed concrete reactor vessel during unrestricted core heatup accidents

The long term core and primary loop heatup of an HTGR subsequent to loss of all forced circulation has been analyzed using a modified version of the CORCON code. The results indicate that if the liner cooling system is operating, or can be restarted within about 60 h, safe cooldown can be achieved, but significant core damage will occur. Without functioning liner cooling system the core heatup will lead to PCRV concrete degradation and the resulting concrete gas releases will ultimately cause containment building failure after 6 to 10 days.     

Steam reformers heated by helium from high temperature reactors

The manifold possibilities of the application of helium-heated steam reformers combined with high temperature nuclear reactors are elucidated in this article. It is shown that the thermodynamic interpretation of the processes does not cause difficulties because of the good heat transfer in helium at high pressure and that helium peak temperatures of 950°C are sufficient for carrying out the process. The mechanical design of the reformer tube does not lead to problems because the helium and process pressures are so chosen as to be approximately equal. The problems of hydrogen and tritium permeation as well as the contamination of the reformer tube with solid fission products seem to be solvable using the knowledge available at present. Furthermore, the various possibilities for the design arrangements of helium-heated reformer tube furnaces are shown. The status of development attained to date is outlined and in conclusion there is a survey regarding the next steps to be taken in steam reformer technology.     

Passive modular gas safety system for a reactor

Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center, RNTs. Translated from Atomnaya Énergiya, Vol. 75, No. 1, pp. 8–13, July, 1993.     

Consequences of delayed air ingress following a depressurization accident in a high temperature reactor

Air ingress is a specific event in a high temperature reactor (HTR). The potential threat posed by air ingress lies in the chemical reaction of oxygen with hot graphite at a temperature above 500 °C leading to reaction heat and graphite corrosion. In order to assess the consequence of air ingress into the reactor, it is postulated that breaks are present above and below the reactor core and that unobstructed ingress of air through them is possible. It is obvious that the air ingress incident has to be preceded by a depressurization accident. For this hypothetical scenario the maximum possible air flow rate through the core resulting solely from the pressure losses in the core is estimated as a function of the break cross-sections exposed above and below the core.In this paper, the thermal behavior of an HTR with prismatic fuel (operating inlet/outlet temperatures 450/850 °C) during air ingress accident conditions has been investigated. In particular, maximum temperatures and burn-off of the fuel and bottom graphite reflector for various air flow rates for the postulated hypothetical scenario have been analyzed. It also indicates the limiting time at which the graphite layer of fuel will be completely burnt-off and the fuel compacts exposed. In addition, the consequences of delayed air ingress after a core heat up following depressurization have been investigated.This paper, thus, throws light on the safety aspects of the new generation HTRs with prismatic fuels (e.g. NGNP/ANTARES) conceived for power generation and process heat application.     

Multiaxial loading tests of high temperature reactor components

This paper provides background information on the theory and tests to develop stress and strain analysis procedures employed in the high temperature component design of the helium cooled reactor prototype plant THTR.     

Mechanical design methods for high temperature reactor components

The work concerned with design codes for high temperature reactor (HTR) components operating at temperatures above 800°C is summarized. Using the experimental results from the German HTR materials development programmes, in particular the time dependent properties, the structural design analysis for an intermediate heat exchanger is discussed, with reference to creep, fatigue, creep buckling and creep ratchetting. The analysis provides the basis for a critical consideration of ASME Code, Case N 47, and the applicability of the code case rules for service temperatures above 800°C.     

Parameter and cost optimizations for a modular stellarator reactor

The physical scaling and cost scaling of a modular stellarator reactor are described. It is shown that configurations based onl=2 are best able to support adequate beta, and physical relationships are derived which enable the geometry and parameters of anl=2 modular stellarator to be defined. A cost scaling for the components of the nuclear island is developed using Starfire (tokamak reactor study) engineering as a basis. It is shown that for minimum cost the stellarator should be of small aspect ratio. For a 4000 MW_th plant, as Starfire, the optimum configuration is a 15 coil, 3 field period,l=2 device with a major radius of 16 m and a plasma minor radius of 2 m; and with a conservative wall loading of 2 MW/m² and an average beta of 3.9%; the estimated cost per kilowatt (electrical) is marginally (7%) greater than Starfire.